Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors

Due to the widespread use of thick-wall Alloy 690 and its corresponding weld metals Alloys 52 and 152 in various replacement, repair, mitigation, and new plant pressurized water reactor (PWR) applications, there is an industry need for an equation or methodology to predict crack growth rates (CGRs) for primary water stress corrosion cracking (PWSCC) of these materials. An international PWSCC CGR Expert Panel was organized by EPRI, with the participation of national laboratories sponsored by the US NRC, to support the development of such PWSCC CGR equations. A database of over 500 Alloy 690 CGR data points and over 130 Alloy 52/152 CGR data points from seven research laboratories was compiled, evaluated and scored for data quality, and assessed to determine the effects of numerous parameters such as temperature, crack-tip stress intensity factor, yield strength, and crack orientation. The process by which these data were evaluated and scored is presented in this paper.

A great deal of PWSCCPrimary Water Stress Corrosion Cracking (PWSCC) testing has been conducted on a range of A690 materials. An expert panel organized by EPRI is working to collect all the available laboratory data for its applicability to actual plant components. This paper will review the considerations that are being used to perform this evaluation, and the results of that evaluation. One of the key variables is cold workCold work , and detailed studies have been conducted to measure the residual strains, so as to determine the amount of cold work that can exist in a heat-affected-zone (HAZ) region of a typical weld. In addition to these measurements on weldments, limitations on bulk cold work on base metal imposed by all vendors will be reviewed. The effect of the orientationOrientation of the specimens tested will be compared with the orientations of typical flaws in operating plants, so as to determine the relevance of the data to those plants.

Extensive SCC growth rate measurements have been performed on Alloy 690 and its weld metals in the past, and this paper focuses on SCC growth rate evaluation of Alloy 52 Alloy 52/152 Alloy 152 welds with a variety of defects and/or weld repairsWeld repairs , and in the dilution zone. Ductility dip cracking dominated the weld defectsWeld defects , and weld repair mockups were fabricated by EPRI Charlotte to be 20% or 50% excavation and repair, as well as welds with a refuse pass every layer. Only low and very low SCC growth rates were observed in all cases. Studies on weld dilution zone effects of varying Cr content were evaluated using welds created with variable ratios of dual-filler-wire feel, which permits definitive SCC growth rate measurements in a homogenous weld without the ambiguity of having the crack front in undefined composition of an actual weld dilution zone.

High-chromium, nickel-based alloys, including alloy 690 base material and alloys 52 and 152 weld filler metals, are used in the primary system of new pressurized water reactors (PWRs), as well as for replacement and mitigation of components in existing reactors. These materials are thought to be highly resistant to primary water stress corrosion cracking (PWSCC), which is observed in plant service for components fabricated from the low-chromium alloys 600, 82, and 182. For over 10 years, the U.S. Nuclear Regulatory Commission has sponsored a laboratory testing program to measure the PWSCC growth rates of alloys 690, 52, and 152 in environmental conditions representative of PWRs, with the intent to support technical bases for the determination of appropriate in-service inspection requirements. In many tests, the low crack growth rates are confirmed. For certain cases, however, such as in highly cold worked alloy 690 and at dilution zones between high-chromium weld metals and low-chromium base metals, PWSCC growth rates are reported to be similar to those observed in alloys 600, 82, and 182. Challenges arise in the use of these data for predictive models given the relatively few numbers of tests performed for some material conditions and uncertainties about the correlation between conditions of test materials and those found in the field. This paper will present perspectives on factors that may be considered for the application of these data to the analysis of plant components.

Recent stress corrosion cracking (SCC)Stress Corrosion Cracking (SCC) crack growth rate (CGR) testing at Argonne National Laboratory (ANL) of Alloy 52/152 weldments near dissimilar metal weld interfaces have found, on occasion, some rather surprisingly high SCC CGRs. Weld overlays of alloys believed to possess superior SCC resistance due to their higher Cr content are typically applied over welds made with SCC-susceptible alloys with the expectation that they will act as a barrier to SCC. However, testing conducted at ANL revealed that the SCC CGRs near the interface between the two welds was in the 10−10 m/s range. Likewise, SCC CGR data in Alloy 152Alloy 152 weld butter near the interface with Low Alloy Steel, which is a region with some dilution of CrCr dilution , found SCC rates as high as 10−10 m/s. In most cases, SCC propagation occurred in a direction perpendicular of that of the dendritic grains—a direction not usually associated with fast SCC propagation. The objective of this paper is to present and discuss the testing results with a focus on the possible paths for fast IG SCC propagation in these weldments.

Stress corrosion crack (SCC) arrest tests have been conducted on composite material specimens to study the SCC susceptibility of SCC resistant materials in hydrogenated deaerated water. SCC tests were performed at 360 °C at a stress intensity factor of ~35 MPa $$ \surd {\text{m}} $$ with ring loadedRing loaded composite material specimens fabricated from SCC resistant materials of Alloy 690, weld metals EN52 and EN625, low alloy steel and stainless steel. Control specimens of Alloy 600 and EN82H were also tested. SCC results and analytical characterization results are discussed.

The mechanism of primary water stress corrosion cracking (PWSCC) in Alloy TT690 was studied from the viewpoint of the influence of hydrogen behavior and cavity formation. Four parameters were examined: cavity formation at crack tips; deformation in the presence of hydrogen gas; dynamic strain effect on mechanical properties; and nickel plating effect on PWSCC propagation. The correlation that was observed between cavity formation at crack tips and crack growth rates indicates the cavities have an important influence on cracking. Results for creep tests done in the presence of hydrogen gas reveal hydrogen lowers the elasticity and makes the alloy brittle at the stress concentrated region. Results for slow strain rate tensile tests indicate that the dynamic strain which accompanies the hydrogen source increases the apparent yield strength. Additionally, nickel plating, which might change corrosion behavior on the alloy surface, suppresses the crack propagation. These observations provide direct support for hydrogen and cavity related mechanism that account for PWSCC of Alloy TT690.

Crack initiation has often been measured using simple, unmonitored tests such as bolt-loaded U bends, where the time to initiation is only estimated by occasional interruption, and the stress drops by ~12% from the change in modulus along with stress relaxation, which can be substantial in some materials and heats. The objective of this study was to develop and demonstrate improved techniques using actively loaded tensile specimens and continuous on-line monitoring of crack development using reversing DC potential drop. To complement the crack initiationCrack initiation data, the crack growth response of the heatsHeat and conditions was also evaluated.

SCC initiation behavior of 15% CF specimens cut from four different alloy 182Alloy 182 weldments was investigated in 360 °C simulated PWR primary water under constant load at the yield stress using direct current potential drop to perform in situ monitoring of SCC initiation time. Within each weldment, one or more specimens underwent SCC initiation within 24 h of reaching full load while some specimens had much longer initiation times, in a few cases exceeding 2500 h. Detailed examinations were conducted on these specimens with a focus on different microstructural features such as preexisting defects, grain orientation and second phases, highlighting an important role of microstructure in crack initiation of alloy 182.

Interactions between multiple stress corrosion cracks (M-SCC) have a major influence on crack growth but are underestimated in models devoted to the evaluation of the lifetime of industrial components. In this study, the growth and interactions between multiple cracks on a sensitized Alloy 600 in a 0.01 M tetrathionate solution, were studied by digital image correlation Digital image correlation (DIC). Cracks exceeding 55 µm in length and 0.45 µm in opening were successfully detected by DIC. The emergence and intensification of interactions modify the growth of the crack colony which evolves from a mostly surface crack propagation (lack of interactions) to in-depth propagation controlled by crack shielding. A multiphysics phase field model was jointly developed and successfully implemented to simulate intergranular M-SCC. It coupled a robust algorithm based on brittle fracture and a diffusion model. The resulting modeling allowed simulating the interactions between cracks and the shielding effects observed experimentally. Finally, 3-D quantification of crack propagation was performed by micro-tomography and digital volume correlation (DVC).

Experiments in hydrogenated steam were performed on several U-bend specimens extracted from two Alloy 82 welds. Results demonstrated that Alloy 82 is susceptible to SCC in hydrogenated steam at 400 °C and its susceptibility depends on its chemical composition, welding process and thermal treatment. The microstructure was characterized in the apex of the U-bend specimens. Chemical analysis were performed by electron probe microanalyser (EPMA) and secondary ion mass spectrometry (SIMS) on several areas in the weld in order to correlate crack initiation with chemical heterogeneities. It was concluded that there are more cracks in the roots of the weld passes where the impurity content (sulfur, titanium and aluminum) is higher.

Cavitation Peening (CP)Cavitation Peening (CP) is achieved by delivery of Ultra-High Pressure (UHP) water through a nozzle, the pressure drop through the orifice creating cavitation bubbles, and delivering these bubbles to the target metal surfaces. Collapse of the bubbles on the surface generates a shock wave resulting in compressive stresses. Complex or regular geometric surfaces can be treated by “coating” them with cavitation bubbles through this simple, robust, and forgiving process. Laboratory tests have demonstrated that the compressive stresses due to UHP CP preventPrimary Water Stress-Corrosion Cracking (PWSCC) PWSCC crack initiation of Alloy 600. UHP CP additionally does not significantly affect roughness and hardness in depth of the mitigated surfaces nor risk to damage the component. UHP CP can be applied cost effectively to the remaining Alloy 600 primary components in order to prevent expensive repairs or replacements. Process parameter development and a tooling qualification program for the application of the UHP CP process on Alloy 600 RPV head penetration nozzles Top head penetration nozzle was completed, and this process was successfully implemented in several PWRsPressurized Water Reactor (PWR) .

Stress corrosion crackingStress corrosion cracking (SCC) initiation tests have been conducted on Alloy 600 in simulated PWRPWR primary water (PW) with the aim of understanding the effect of surface condition produced by different machining methods on the time to initiation. Surface roughness, residual stress and cold work were characterised using profilometry, X-ray diffraction (XRD), micro-hardness and electron backscatter diffraction (EBSD). Initiation tests used tensile, button-headed specimens manufactured from 15% cold rolled, low temperature annealed Alloy 600 to relate machining parameters to PWSCC initiation susceptibility. Tests were conducted at 360 °C with active loading corresponding to 1% plastic strain. The onset of initiation was detected by in situ direct current potential drop (DCPD) measurement. Results indicate that residual stress and orientation of the crack plane in relation to cold work are critical parameters for Alloy 600 PWSCC initiation susceptibility. This suggests that knowledge of surface roughness cannot be used as the sole acceptance criteria for surface condition with regard to PWSCC susceptibility.

SCC initiation results in simulated PWR primary have been obtained on one mill annealed alloy 600Alloy 600 plate heat in the cold-worked condition. Twelve specimens with similar cold workCold work levels were tested at constant load and three showed much shorter SCC initiation times (<400 h) than the nine others (>1200 h). Post-test examinations revealed that these three specimens all feature an inhomogeneous microstructure where the primary crack always nucleated along the boundary of large elongated grainsLarge elongated grain protruding normally into the gauge. In contrast, such microstructure was either not observed or did not extend deep enough into the gauge in the other specimens exhibiting ~3–7× longer initiation times. In order to better understand the role of this microstructural inhomogeneity in SCC initiation, high-resolution microscopy was performed to compare carbide morphology and strain distribution between the long grains and normal grains, and their potential effects on SCC initiation are discussed in this paper.

Alloy 690 is a nickel-based alloy (60% Ni, 30% Cr, 10% Fe) used in nuclear Pressurized Water Reactors (PWR) for different components and welds (steam generator tubes etc.). They are subjected to thermal ageing up to 60 years which could lead to an order-disorder transition (Ni2Cr ordered phase formation) by a diffusion-assisted mechanism. This transformation might modify mechanical properties and is suspected to influence the stress corrosion resistance of the affected components. To study ordering kinetics, hardness, thermoelectric power (TEP)Thermoelectric power alongside transmission electron microscope (TEM) observations were conducted on Ni-33%Cr alloys with different iron contentsIron content (0–3 wt%) after various ageing thermal treatments. The ordering activation energies have been determined: they are found to be independent of the iron content. A correlation between macroscopic properties and TEM diffraction results is proposed. Finally, the distribution of iron between matrix and ordered domains was studied.

Mechanical property degradation due to isothermal ageing is of potential concern for alloys based on the Ni–Cr binary system, such as Alloys 625 and 690. The disorder-order phase transformation, which is the primary source of embrittlement, has been studied in Ni–Cr model alloys by experimental approaches. Model alloys with different stoichiometries have been isothermally aged up to 5000 h at three temperatures (373, 418, and 475 °C) and characterized via nano-indentation, atom probe tomography, and transmission electron microscopy. Results show that off-stoichiometry alloys exhibit ordering but at a slower rate than stoichiometric (Ni/Cr = 2.0) alloy.

The objective of this work is to evaluate the effect of hardening on the susceptibility of a Ni-30.7Cr wt% (Ni-33 at.%Cr) model alloy to stress corrosion cracking (SCC)Stress corrosion cracking and to low temperature crack propagation (LTCP)Low temperature crack propagation . Unlike previous studies that employ cold work to induce susceptibility to SCC, this work utilizes isothermal ageing to produce the long range orderedLong range order Ni2Cr phase. Samples were aged at 475 °C for durations up to 3176 h in order to produce hardness values between 70-100 HRB. After ageing, precracked compact tension specimens were tested for susceptibility to primary water SCC and to LTCP. Samples aged for less than 200 h (70-80 HRB) showed very high resistance to SCC, while intergranular cracking was observed in samples aged for longer times. An activation energy of 129.7 ± 17.6 kJ/mol and a yield strength exponent of 6.4 ± 2.1 were measured for SCC growth at constant KI and ΔEcP conditions, consistent with observations for Alloy 690Alloy 690. Hardening via long range order had no measureable effect on the toughness of the alloy in air, but degraded the toughness and promoted intergranular fracture in hydrogen deaerated water (i.e. it caused susceptibility to LTCP). The similarity of the yield strength dependence to cold worked Alloy 690 and the common temperature dependence (~130 kJ/mol) to A600, X-750, A690, etc. suggests a common SCC mechanism for all these alloys. Hardening via long range order is a novel method to induce SCC susceptibility in Ni-30 wt%Cr alloys, which avoids some microstructural damage, inhomogeneity, and orientation effects that complicate testing of cold worked material.

Thermally aged nickel based AlloyNickel based alloy 600 was investigated to evaluate the effects of long-term thermal aging and triaxial stress on primary water stress corrosion crackPrimary water stress corrosion cracking initiation behavior. Long-term thermal aging was simulated by heat treatment at 400 °C, a temperature that does not cause excessive formation of second phases that cannot form in nuclear power plant service conditions. Triaxial stress was applied by a round notch in the gauge length of some test specimen; other specimens were smooth. Slow strain rate tests (SSRT) monitored by the direct current potential drop method were conducted to evaluate stress corrosion crack initiation susceptibility of the thermally aged specimens in the primary water environment. For smooth specimens (which experience uniaxial stress), the susceptibility of those thermally aged for the equivalent of 10-years was the highest, while the susceptibility of the as-received specimens was the lowest. However, for the notched specimens (which experience triaxial stress), the specimens thermally aged for the equivalent of 20-years showed the highest susceptibility, while the as-received specimens showed the lowest.

Alloy 718 is an age hardenable, nickel-base alloy used in fuel assembly of Pressurized Water Reactors (PWRs) by virtue of its high strength and resistance to corrosion and stress corrosion cracking (SCC). SCC susceptibility is affected by the microstructure developed during thermal mechanical treatments. The SCC behavior of alloy 718 in three different thermal mechanical treatments (TMTs) and two different heats was studied in PWR primary water environment using constant extension rate tensile (CERT) tests. TMTs have a significant effect on the microstructure and thus the mechanical behavior and the SCC susceptibility of alloy 718. TMTs using a solution anneal at 1093 °C with a two-step ageing treatment (1093 °C/1 h + 718 °C/8 h + 621 °C/8 h) exhibited the best SCC resistance.

A series of Alloy 718 specimens were irradiated in the Halden Reactor under mechanical tensile stresses and in a chemical environment and temperature representative of Pressurized Water Reactor service conditions. The specimens were miniature pin-loaded dogbones, heat treated using either a direct aging cycle or the same aging heat treatment preceded by a solution anneal. Applied stresses ranged between 920 and 1200 MPa. Fracture surfaces examined by SEM displayed a mixture of intergranular regions perpendicular to the applied stress and smoother regions at various angles to the applied stress. It is concluded that intergranular cracking proceeded until the stress on the remaining ligament was sufficient to cause prompt ductile fracture. Fluence at fracture occurred over a range of seven orders of magnitude, with no correlation to applied stress. Time at fracture spanned a much smaller range and was broadly, though weakly, inversely correlated with stress. It appears that time in the environment is a better predictor of failure than is fluence.

Thermal ageing promotes intergranular carbide precipitation and atomic ordering reaction in most commercial nickel-base alloys, and it affects the long-term primary water stress corrosion cracking (PWSCC) resistance of pressurized water reactor components. Alloy 690 with 9.8 wt% Fe was solution annealed and heat-treated at low temperature, then aged between 350 and 550 °C for 10,000 h. No direct observation of ordering was possible, but variations in hardness and lattice parameter suggested the formation of short-range ordering (SRO) with a peak level upon ageing at 420 °C, while a disordering reaction occurred at higher temperatures. Heat treatment induced ordering before thermal ageing was compared to the solution-annealed state. Thermal ageing resulted in the precipitation of Cr-rich M23C6 carbides at grain boundaries and twin boundaries. Although no link between SRO and an increase in strain localization was observed, the combination of intergranular carbide formation and SRO over longer ageing times was deemed detrimental to the PWSCC resistance of Alloy 690 TT.

Two mechanisms are studied to explain the asymmetrical chromium depletionsChromium depletion observed ahead of SCC crack tips in nickel-base alloys: diffusion-induced grain boundary migration (DIGM)DIGM and plasticity-enhanced diffusionDiffusion . On the one hand, DIGM is evidenced in a model Alloy 600Alloy 600 by focused ion beam (FIB) coupled with scanning electron microscopy (SEM) cross-section imaging and analytical transmission electron microscopy (TEM) after annealing at 500 °C under vacuum and at 340 °C after exposure to primary water. The occurrence of grain boundary migration depends on the grain boundary character and misorientation. On the other hand, the effect of plasticityPlasticity on chromium diffusion in nickel single-crystals is investigated by performing diffusion tests during creep tests at 500 and 350 °C. An enhancement of Cr diffusion is observed and a linear relationship between the diffusion coefficient and strain rate is evidenced. At last, in an attempt to discriminate the two mechanisms, an analytical modeling of the Cr-depleted areas observed at propagating SCC crack tips is proposed.

It is well established that thermally treated (TT) alloy 600 exhibits superior resistance to intergranular stress corrosion cracking (IGSCC) in pressurized water reactor (PWR) primary water environments than do solution annealed (SA) and mill annealed (MA) equivalents of the same material. This improved resistance is nominally ascribed to the prevalence of grain boundary (GB) Cr-carbide precipitates (M7C3 or M23C6). In this study, we perform high-resolution characterization by scanning transmission electron microscopy (STEM) and atom probe tomography (APT) on a heat of alloy 600 with a very low concentration of C (0.01 at.%) and a modest B concentration (46 appm). After SA+TT annealing, this heat exhibits IG Cr-boride precipitates (Cr5B3) in the absence of Cr carbides. Despite IG precipitation of Cr5B3, the pristine GB shows no measurable Cr depletion and B remains segregated at the GB. During intergranular attack (IGA) in PWR primary water at 360 °C, the Cr5B3 precipitates dissolve rapidly. The observed IGA is subtly different from what is typical for alloy 600, specifically with higher concentrations of Li and B within the corrosion oxides while the GB ahead exhibits less depletion of oxidizing species (e.g. Cr, Fe, Si and B) than other alloy 600 heats. Together these observations suggest that IG precipitation of Cr5B3 in the absence of Cr carbides has a neutral to slightly positive effect on the IG corrosion resistance of alloy 600 GBs in PWR primary water.

Advanced electron microscopy and surface science techniques were applied to characterize inter- and intragranular oxidation in Ni–Fe–Cr alloys after exposure to 480 °C hydrogenated steam. Intragranular internal Fe and Cr oxidation was observed in all cases while intergranular oxidation, exclusively external or penetrative, varied depending on the Cr content of the alloy. The kinetics and morphology of intragranular internal oxidation and nodule growth were studied through successive short-term exposures with characterization performed between exposures. FIB 3D serial sectioning was used to reconstruct volumes containing oxidized grain boundaries and revealed that diffusion-induced grain boundary migration may play a fundamental role in increasing the outward flux of Cr, Ti, and Al near grain boundaries, depending on the extent of intergranular Cr carbide precipitation. In addition, atom probe tomography was used to study the behaviour of minor impurity elements, Al and Ti, and initial oxidation processes. Further analyses of oxidized samples using three-dimensional ToF-SIMS are also discussed.

Oxidation studies were performed on Alloy 600 in situ in high-temperature H2/N2 + H2O vapor mixture at 400 °C. The initial stages of preferential intergranular oxidation (PIO), shown to be an important precursor phenomenon for primary water stress corrosion cracking, have been successfully identified using the in situ approach. The behaviour of Alloy 600 was observed in real time using the Protochips environmental cell and analysed via analytical electron microscopyAnalytical electron microscopy (AEM). Post in situ AEM analyses were compared with previous ex situ post-exposure characterization results obtained from bulk specimens, demonstrating good agreement. The in situ results confirmed the grain boundary migration and intergranular oxide formation in solution-annealed Alloy 600. The excellent agreement between the in situ and previous studies demonstrates that this approach can be used to investigate the initial stages of PIO relevant to nuclear power systems.

Low pressure superheated H2-steam system has been extensively used in the past years to accelerate the oxidation kinetics while keeping the conditions representative to PWR primary water. One of the most important requirements of this environment is that needs to replicate the Ni/NiO transition. However, despite several studies have been carried out by different research groups in H2-steam environment, there is still some level of uncertainty over the thermodynamic of the oxidation process. In this study, the Ni/NiO transition in hydrogenated steam was investigated via electrochemical potential measurements using a Ni/NiO solid state reference electrode. Furthermore, solution annealed Alloy 600 coupons were exposed to H2-steam at 480 ℃ in order to examine the effect of oxidizing conditions with respect to the Ni/NiO transition on the preferential intergranular oxidation. The effect of the redox potential on the preferential intergranular oxidation is discussed in the context of the precursor stages of stress corrosion cracking for Alloy 600.

The effect of dissolved hydrogen (DH)Dissolved hydrogen on primary water stress corrosion crackingPWSCC of nickel base alloys has been of intense interest for plant operators worldwide. In this study, crack growth rates of Alloy 600Alloy 600 were measured in simulated PWR primary coolant at 330 °C with DH levels of 5, 16, 45 and 75 cc H2/kg H2O, respectively. The oxide films formed in the crack tip regions were examined using transmission electron microscopy (TEM). The results show low and similar crack growth rates at all DH levels, without a maximum at 16 cc H2/kg H2O. The low DH content favors nickel oxide formation at the crack tip region, whereas the high DH level favors Me3O4 type spinel formation. Also, the oxide films were found to grow epitaxially on some metal grain surfaces in the cracks. The possible effects of alloy composition on the oxide films formed, and the effect of DH on the crack growth are briefly discussed.

Stress corrosion cracking (SCC) in Alloy 600 has been studied in simulated pressurized water reactor (PWR) primary water at various temperatures. A clear correlation between temperature and crack growth rate (CGR) was found showing that the CGR increased monotonously within the range of temperatures used in this study (320–360 °C). In order to understand the temperature dependence of CGR, high-resolution characterization was used to study the crack tips. The crack tips obtained from different temperatures were analyzed by high-resolution analytical transmission electron microscopy (TEM) to reveal the crack tip morphology and chemistry, which enable the study of a thermally activated diffusion-based mechanism operating during SCC propagation. Transmission Kikuchi diffraction (TKD) was used to investigate mechanical response-based mechanisms in SCC propagation through quantifying the size and extent of plastic deformation around the crack tips. Results obtained in this study show that the thermally activated diffusion along the grain boundary increased with temperature while the changes of plastic deformation around the crack tip were small and nearly independent of temperature, suggesting that a thermally activated diffusion-based mechanism was dominant.

Long-term grain boundary (GB) damage evolution and stress corrosion crack initiation in alloy 690 are being investigated by constant load tensile testing in high-temperature, simulated PWR primary water. Six commercial alloy 690 heats are being tested in various cold work conditions loaded at their yield stress. This paper reviews the basic test approach and detailed characterizations performed on selected specimens after an exposure time of ~1 year. Intergranular crack nucleation was observed under constant stress in certain highly cold-worked (CW) alloy 690 heats and was found to be associated with the formation of GB cavities. Somewhat surprisingly, the heats most susceptible to cavity formation and crack nucleation were thermally treated materials with most uniform coverage of small GB carbides. Microstructure, % cold work and applied stress comparisons are made among the alloy 690 heats to better understand the factors influencing GB cavity formation and crack initiation.

Long-term constant load stress corrosion cracking (SCC) testing for alloys 690/152/52 at 360 ℃ is ongoing, showing no rupture for more than 105 h, suggesting immunity to primary water (PW) SCC initiation under stress level assumed for primary circuit components in pressurized water reactor (PWR) plants. Since the mechanical plug for steam generators has the largest cold work strain, a mock-up PWSCC test, using a mechanical plug of alloy 690, was also performed for evaluation of time to failure under stress and cold work conditions assumed for operating plan. As a result, it was proven that no crack initiated up to approximately 4 × 104 h at 360 ℃. PWSCC susceptibility was also evaluated in terms of crack growth rate (CGR). The CGR of alloy 690 increased after cold working, and the degree of increment is significantly affected by the nature of carbide precipitate along grain boundaries. It was found that increase in CGR caused by cold working remained relatively low when grain boundary carbides precipitated continuously along grain boundaries and coherently with the matrix. Contrarily, CGR grew higher in the materials with lower coherency. It was also revealed that alloy 690 with no grain boundary carbides (solution annealed alloy) showed a small increase of CGR after cold working.

The effects of local strain distribution on primary water stress corrosion cracking of cold-rolled Alloy 690 with an inhomogeneous microstructure were investigated by measuring dislocation densities using transmission electron microscopy. Many intragranular carbides with a Cr-rich M23C6 structure were dispersed in the fine grains. The results showed that the dislocation densities near the intragranular carbides were high, regardless of the degree of cold-rolling. Below 20% cold-rolling, the dislocation densities near grain boundaries were higher than that in the grain interior. Meanwhile, the dislocation densities in the grain interior increased to similar value of the grain boundary with increasing degree of cold-rolling up to 40%. The results indicate that the intragranular carbides dispersed in the fine grains play an important role in the local strain distribution of a cold-rolled Alloy 690 with an inhomogeneous microstructure. It is suggested that the high local strain in the grain interior in a severely cold-rolled Alloy 690 induced by interaction between dislocation and intragranular carbides could be responsible for the mixed cracking mode and the high crack growth rate.

The evolution of carbides precipitated on grain boundaries near triple junctions in Inconel Alloy 690 following aging treatments at 715 °C for 2, 15 and 50 h were investigated by SEM and EBSD. The results show that, aging time does not influence the morphology of the carbides precipitated at grain boundaries, however, it does influence the morphology of carbides precipitated at triple junctions. Tiny carbides are precipitated at coherent Σ3 grain boundary, bar like carbides are precipitated at both sides of incoherent Σ3 grain boundary, while on only one side of Σ9 grain boundary, and the morphology of carbides precipitated on Σ27 and random grain boundaries are similar. Therefore, the type of nearby grain boundaries can influence the morphology of carbides precipitated at the other grain boundaries.

The Stress Corrosion Crack Growth Rate (SCCGR) in heavily cold rolled Thermally Treated (TT) Alloy 690 exposed to simulated PWR primary water at 360 °C increases with increasing cold rolling ratio, but the SCCGR in cold rolled Mill Annealed (MA) Alloy 690 remains very low, regardless of cold rolling ratio. Cavities were detected near GB carbides in heavily cold rolled TT Alloy 690 before the SCC tests. There is a good correlation between the existence of cavities near GB carbides and high SCCGRs in heavily cold rolled TT Alloy 690. The number of cavities increases with increasing cold rolling ratio and is affected by heating in air at 400 or 475 °C for ~2000 h and by exposure in simulated PWR primary water at 360 °C. However, the cavities were detected not only in the stressed area but also in the stress-free area of the SCC test specimens of heavily cold rolled TT Alloy 690. By contrast, the effect of Ni content on SCCGRs in Ni base (25–30%)-Cr-Fe alloys is not significant for similar amounts of GB carbide precipitation. The high SCCGRs in heavily cold rolled TT Alloy 690 may be caused by a high density of lattice defects, cavities near GB carbides, cracking of M23C6 primary GB carbides, and hydrogen absorption, but there is no possibility of creep damage at the test temperature of 360 °C. More detailed tests will be needed to confirm this hypothesis.

This study was aimed at investigating the intergranular attack near a stress corrosion crack (SCC)SCC of alloy 690 in simulated pressurized water reactor (PWR)PWR primary water environment. Solution annealed alloy 690Alloy 690 was evaluated for its SCC initiation susceptibility in 360 °C hydrogenated pure water using slow strain rate tensile technique. After the test, a grain boundary showing SCC initiation was sampled with Focused Ion Beam (FIB) milling. The microstructure and elemental distribution near the crack tip were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). The results show that intergranular oxidationIntergranular oxidation occurs ahead of the crack tip and is preceded by diffusion induced grain boundary migrationBoundary migration . The oxides at the crack tip are mainly composed of NiO and Cr2O3 which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr2O3 dominates at the oxide/substrate interfaces and the very tip region.

Components of reactor core internals suffer from Irradiation Assisted Stress Corrosion Cracking. Here is studied 08Ch18N10T austenitic stainless steel acquired from decommissioned NPP Nord Unit 1, VVER 440-230 type in Greifswald, where had been irradiated to 5.2 dpa. The material was tensile tested at 20 °C in air and slow strain rate tested at 320 °C in air and in water. SEM observations of the fracture surface found ductile fracture for the air tests, but areas of intergranular fracture typical of IASCC in the water. This paper emphasizes the microscopic examination from three samples to determine the underlying physical damage processes. TEM observations close to the fractured surface focused to the interaction of dislocations with local radiation damage defects and grain boundaries owing to different test conditions. Determination of local chemical composition around the grain boundaries indicated radiation induced segregation; as well presence of helium gas in voids. The observation of tensile tests found the presence of twinning and regions of strained martensite transformation. The nano features of tests at elevated temperature were tangled dislocations, similar in air and water. No effect of the water environment on the deformation structures was observed.

5 meV Ni++ and Fe++ ion irradiations were performed to investigate radiation-induced precipitates evolution in a cold-worked 316 austenitic stainless steel at high doses and temperatures. The irradiation conditions were 23 dpa at 380 °C, 130 dpa at 380 °C, 23 dpa at 500 °C, and 15 dpa at 600 °C. TEM selected electron diffraction (SAED), TEMTEM dark-field imaging and energy dispersive spectroscopy (EDS) Energy Dispersive Spectroscopy (EDS) mapping were used as complementary techniques to determine crystallography, morphology and chemical composition of radiation-induced precipitates. The precipitates were predominantly in form of the Ni–Si rich γ′ phase at all irradiation conditions. The EDS analysis further determined Ni–Si–Mo–P and Ni–Si–Mn rich precipitates after irradiation at 380 and 600 °C, respectively. The precipitates were found close to saturated state between 23 and 130 dpa at 380 °C irradiation conditions. A different effect of higher irradiation temperatures was found between 500 and 600 °C. In case of the irradiation to 23 dpa at 500 °C, the average size of precipitates was similar to irradiations at 380 °C, but the density was lower. However, the precipitates revealed large size and very low density following the irradiation to 15 dpa at 600 °C. The original dislocation network introduced by cold-working was found as dominant sink for intra-granular solute radiation-induced segregation (RIS) and possibly took place as primary nucleation site of radiation-induced precipitates at irradiation temperatures 380 and 500 °C. At the temperature 600 °C, the RIS at dislocation network almost vanished and the main nucleation sites became twin boundaries as more energetically favorable intra-granular sinks.

Understanding the changes in mechanical properties and failure mechanisms as a function of radiation damage is important for long-term operation of structural components in nuclear reactors. Due to the expense, the activation of the sample, and the long duration of neutron irradiation, ion beam irradiations (proton-irradiation and, increasingly, heavy-ion irradiation) are used as surrogates for neutron irradiation. However, the shallow irradiation depths of ion-beam irradiation have restricted mechanical property measurements until the recent advent of small-scale mechanical testing. In previous studies, nano-hardness and yield strength of proton-irradiated 304SS were measured using nanoindentation and in situ microcompression respectively. This study develops an in situ microtensile testing method to provide direct stress-strain curves, including the strain to failure, which previous studies do not provide. In addition, a novel way of quantifying irradiation-induced susceptibility to slip band formation in microscale specimens was demonstrated. Lastly, the paper introduces a new technique for measuring grain boundary strength, demonstrated on an oxidized grain boundary of Alloy 600 exposed to primary water chemistryPrimary water chemistry environment.

The next generation of light water reactors, resource renewable BWR (RBWR), which can be burned trans uranium (TRU) is currently under development at Hitachi. The RBWR requires a high flux of fast neutron for efficient burning of the TRU, which is four times as large as that of the ABWR. Therefore, structural materials require both a high resistance to corrosion and to irradiation. In this study, oxide dispersion strengthened austenitic Stainless steel stainless steelsODS (ODS-ASUS) with high corrosion resistance have been developed. The objective of this research is to evaluate irradiation resistance and SCC Stress Corrosion Cracking (SCC) susceptibility in a simulated reactor water environment for the ODS-ASUS. The materials were irradiated with 6.4 MeV Fe3+ at 673 K up to 8.0 dpa using the DuET facility at Kyoto University. The creviced bent beam (CBB) test is conducted to assess the SCC susceptibility in the hot water, 288 °C, 8 MPa with a dissolved oxygen of 8 ppm.

This paper discusses applications of spherical nanoindentation stress-strainNanoindentation stress-strain curves in characterizing the local mechanical behavior of materials with modified surfaces. Using ion-irradiated tungsten as a specific example, this paper demonstrates that a simple variation of the indenter size (radius) can identify the depth of the radiation-induced-damage zone, as well as quantify the behavior of the damaged zone itself. Using corresponding local structure information from electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM) we look at (a) the elastic response, elasto-plastic transition, and onset of plasticity in ion-irradiated tungsten, zirconium and 304 stainless steel under indentation, and compare their relative mechanical behavior to the unirradiated state, (b) correlating these changes to the different grain orientations as a function of (c) irradiation from different sources (such as He, W, and He+W for tungsten samples).

Stress corrosion cracking (SCC) has been found on surfaces of reactor internals which are heavily cold-worked by machining. When repair welding is applied to the cracked region, the cold-worked surface will be included in the welding region. Therefore, the effect of the cold-worked layer on the weldability of irradiated stainless steels (SSs) needs to be clarified. In this study, helium bubblesHelium bubble formed by welding on irradiated SSs were investigated by measurements and comparisons to published results. TIG weldingWelding was performed on an irradiated SS plate with a cold-worked layer generated by wire peening. Re-crystallization was confirmed in the weld after peening at the surface near the weld heat affected zone (HAZ). The helium bubble number density along the grain boundary in the re-crystallized region was the same as in the non-re-crystallized region, whereas the bubble diameter was smaller. Helium bubble growth was suppressed by the re-crystallization behavior.

PWR structural components built from austenitic steels accumulate large amounts of helium and hydrogen during service and it is desired to predict the concentrations of these gases not only for higher neutron exposures, but also for subsequent ion-induced extension of the damage dose. While the neutron-induced sources of helium are well-known, there is often uncertainty in the local thermal neutron fluence, which constitutes the major uncertainty for prediction of helium production. Combining earlier measurements of helium at lower dpa levels with knowledge of the 59Ni behavior we can extrapolate to higher dpa levels since the helium concentration is an excellent retrospective dosimeter of the thermal fluence. Predictions of hydrogen are more difficult, however, due to helium-nucleated cavities storing both transmutant and environmental hydrogen. Subsequent ion irradiation of neutron-irradiated material can then use double-ion and triple-ion injection to maintain the neutron-relevant gas cogeneration rates.

Self-ion irradiation is currently being used to explore the relative void swelling resistance of various candidate advanced alloys for a wide variety of nuclear systems, including light water reactor (LWR) reactors. The credibility of using this surrogate irradiation technique to evaluate potential in-reactor behavior requires that certain facets of neutron-induced behavior be reproduced in the ion simulation. Of particular importance is the ability of ion irradiation to produce the anticipated post-transient swelling rates for fcc and bcc iron-base alloys characteristic of reactor irradiation at ~1.0%/dpa and ~0.2%/dpa, respectively. Using a model duplex Fe-9Cr-C alloy irradiated at 450 ℃ with 8 MeV Fe+ ions it is shown in this study that a post-transient rate of ~0.2%/dpa is observed in the ferrite phase after an incubation period of ~60 dpa. It is also shown that the ferrite phase attains this rate first, while the tempered martensite phase exhibits a longer transient delay prior to the onset of high-rate swelling.

Neutron-induced swelling in austenitic and ferritic steels is sensitive to the carbon level in the steel, as well as its distribution in matrix or precipitates. It has recently become known that ion-irradiation to high dpa levels leads to a progressive ion-beam-induced increase in carbon concentration and precipitation within the ion range, with concurrent reductions in void swelling. This neutron-atypical phenomenon imperils the credibility of ion simulation for light water reactor applications. A series of experiments involving pure iron and a structural alloy HT9, were conducted to identify the source and distribution of injected carbon. It was found that negatively-charged carbon atoms are entrained in the self-ion beam by a Coulomb drag effect, and thereby delivered at low drift energy to the irradiated surface, followed by ion-beam-mixing and diffusion. A technique for filtering out contaminants, especially carbon, oxygen and nitrogen, was developed and resulted in higher, more neutron-relevant swelling levels than achieved without filtering.

Most of the available void swelling (VS) Void Swelling (VS) data are from fast reactors rather than PWRs, but there are indications that the saturation VS rate for PWR-relevant conditions is at least an order of magnitude smaller than that observed in fast reactors. In 2005, VS screening criteria (temperatures ≥320 °C (608 °F) and fluence ≥20 dpa) were calculated. Since publication of these initial screening criteria, a physics-based model has been developed for the prediction of VS in irradiated austenitic stainless steelStainless steel components. Comparisons between experimental data derived from density measurements and transmission electron microscopy (TEM) characterizations suggest that the Cluster DynamicsCluster Dynamics model is capable of predicting the evolution of the irradiated microstructure under PWR conditions.Pressurized Water Reactors (PWR) The Cluster Dynamics model estimates <1.5% VS for solution-annealed or cold-worked austenitic stainless steel at temperatures below 320 °C (608 °F) and doses <20 dpa for all displacement rates. Therefore, the previous conservative screening criteria originally calculated in 2005 are retained.

A mean-field, cluster dynamics model of the microstructure evolution in austenitic steels of light water reactors which reproduces the incubation period of swelling has been developed for the first time. In agreement with observations, it predicts that, although the void nucleation starts from the very beginning of irradiation, their growth beyond certain, relatively small size is delayed until the onset of the transient period of swelling. Such a delayed growth of voids is explained by the frequent interaction of voids randomly distributed over the volume with one-dimensionally migrating clusters of self-interstitial atoms, which are produced in cascades of atomic displacements. The incubation period of swelling is followed by the transient stage, when voids start to grow with increasing rate due to development of the experimentally-observed spatial correlations between voids and extended defects, such as second-phase precipitates and dislocations, which screen voids from the mobile clusters. A critical role of residual gas on void nucleation, which diminishes importance of He atoms from transmutation reactions, is revealed.

The effects of irradiation on Inconel® (Inconel is a registered trademark of Special Metals Corporation and its subsidiaries) X-750, have been shown to lead to embrittlement and intergranular fracture. This is now widely accepted to be a result of intergranular helium bubbles over the fluence range studied. This paper provides a quantitative assessment and a detailed discussion of the radiation-induced defects including; helium bubbles (size and density distribution, and grain boundary area fraction), dislocation loops and stacking fault tetrahedra, and the disordering and dissolution of secondary gamma prime precipitates. The microstructural evolution will be presented and discussed as a function of dose (from ~5.5 to ~80 dpa), helium concentration (~1300 to ~25,000 appm helium), and irradiation temperature (~120–280 to ~300–330 °C).

A novel lift-out, push-to-pull, micro-tensileMicro-tensile , small scale mechanical testing (SSMT)Small Scale Mechanical Testing (SSMT) technique was developed to assess the yield strength, failure strength, and failure mechanisms of activated ex-service Inconel X-750 removed from the CANDU nuclear reactor core after extended service. Neutron irradiated Inconel X-750 components fail in an intergranular manner. Because these ex-service components are less than 1 mm in thickness, conventional tensile specimens cannot be fabricated from them. Thus, large-scale testing is not possible, and specimens on the order of 1 μm × 1 μm × 2.5 μm (thickness × width × gauge length) containing individual boundaries were fabricated in order to assess the grain boundary strengthGrain boundary strength of the material as a function of irradiation temperature and dose. The variability introduced by differences in thermo-mechanical processing during fabrication was also assessed. Application of this new Micromechanics micro-tensile testing technique to non-irradiated Inconel X-750 gives good agreement with the bulk yield strength of the nickel superalloy, 1070 MPa. From SSMTs, the measured yield strengths of non-irradiated specimens were 1001 MPa at the outer edge and 1043 MPa at the center of the component. Cold work, introduced by grinding of the outside surface of the component, reduces ductility, as does irradiation. Initial tests indicate that away from the surface in the center, the boundary strength was reduced by ~456 MPa after irradiation to 78 dpa at an average irradiation temperature of 180 °C; the corresponding ductility decreased from 16.6 to ≤2.3% total elongation. Testing is a work in progress and more tests are needed for higher precision with regards to grain boundary strength reduction.

Proton irradiation is a well-known useful experimental technique to study neutron irradiation-induced phenomena in reactor core materials. Type 316 austenitic stainless steel was irradiated with 2 meV protons to doses up to 10 displacement per atom at 360 °C, and the various effects of the proton irradiation on the microstructural changes were characterized with transmission electron microscopy and atom probe tomography. Typical irradiation damage mainly consisted of small dislocation loops, cavities, tiny precipitates and network dislocations. Ni and Si were enriched, whereas Cr, Mn and Mo were depleted on the grain boundaries associated with irradiation-induced segregation. Ni–Si rich clusters were also found in the matrix. A new method to prepare TEM specimens of a proton-irradiated material is suggested, which was shown to be a relatively simple and effective method to chemically eliminate the inherent surface damage induced by a conventional high-energy focused ion beam and subsequent low-energy ion milling treatments.

Although service experience of austenitic stainless steels exposed to PWR primary coolant has been good, stress corrosion crack propagation has been observed in laboratory tests in the presence of ≥15% cold work. Data on crack initiation are much more limited and this study therefore aims to improve the understanding of the conditions under which crack initiation and subsequent development of stress corrosion crackingStress Corrosion Cracking (SCC) might be possible. Testing was performed on two heats of Type 304/304L stainless steelStainless steel under slow strain rate tensile loading. A range of analytical techniques were used to characterize the resultant precursor features and cracking, and digital image correlation before and after testing was also used to evaluate the influence of localized deformation on SCC. The results indicate that crack initiation can occur in austenitic stainless steels exposed to good quality primary coolant under dynamic straining conditions; additional testing underway under more plant-representative conditions will be reported later. Significant influences of steel microstructure on crack initiation susceptibility were observed.

Slow strain rate tensile (SSRT) tests Slow Strain Rate Test (SSRT) were conducted on conventional and tapered samples manufactured from forged Type 304L stainless steel304L Stainless steel to assess the stress corrosion cracking (SCC) Stress Corrosion Cracking (SCC) initiation behaviour in simulated PWR primary water. Several testing and microstructural parameters were investigated in order to explore the conditions under which crack initiation might occur. Surface preparation appeared to play a very important role on SCC initiation whereby the machined surfaces were the least susceptible to SCC initiation whilst oxide polishing suspension (OPS) polished surfaces were more susceptible. On the machined surfaces the cracks were always transgranular (TG) in nature and associated with the machining marks. Conversely, on fine polished surfaces with oxide polishing suspension the crack morphology was mainly intergranular in nature, although minor transgranular cracking was observed. The regions in the proximity of the δ-ferrite/austenite interface were shown to be very susceptible to SCC initiation especially on the OPS polished surfaces and this was attributed to the strain localization upon dynamic deformation. Furthermore, intragranular inclusions appeared to dissolve and act as initiation sites for transgranular cracking to occur. The roles of strain rate, dynamic deformation and microstructure on the initiation of SCC are also discussed.

Annealed and cold-worked stainless steelStainless steel 316L samples with machined and polished surfaces were tested in simulated pressurized water reactor (PWR) primary water under slow strain rate tensile (SSRT) test conditions to investigate stress corrosion cracking (SCC)Stress Corrosion Cracking (SCC) initiation. Roughness, residual stress and cross-sectional microstructure of the as-machined samples were characterized before SSRT tests. Plan view and cross-sectional examinations were performed after the test. Pre-test characterization indicated that a deformation layer was present on the machined surfaces. This deformation layer consisted of an ultrafine-grainedUltrafine grain layer on the top and deformation bands underneath. The thickness of the deformation layer on the annealed material was greater than that on the cold-worked material. Post-test characterization revealed that the SCC initiation behaviors of the as-machined and polished surfaces were different for both annealed and cold-worked materials. MachiningMachining increased SCC initiation susceptibility of the annealed material as many shallow cracks initiated along the machining marks in the machined surface, and it decreased the SCC initiation susceptibility of the cold-worked material as a reduced number of cracks were identified in the machined surface compared to the polished surface. The factors influencing SCC initiation are also discussed.

Initiation and propagation of cracks under simulated primary water conditions and different slow strain ratesSlow Strain Rate Test (SSRT) have been studied for an austenitic 304-type stainless steel. Two surface finishes were used to better understand the conditions that trigger stress corrosion cracking (SCC)Stress Corrosion Cracking (SCC) . The main objective is to identify the mechanism(s) that govern the initiation and propagation of SCC and the influence of microstructure. Crack morphology, stress localisation and local chemical composition were characterized for all samples studied. The characterization methodology includes scanning electron microscopy (SEM), 3D focused ion beam (FIB), Transmission Kikuchi Diffraction (TKD)Transmission Kikuchi Diffraction (TKD) , and analytical scanning transmission electron microscopy (STEM).

The field experience of austenitic stainless steels in PWR primary circuits has been generally outstanding. However, a review made by EPRI indicates that significant SCC cases occurred in low flow or stagnant zones where the primary water was probably contaminated by trapped oxygen and/or radiolysis products. In addition, the deleterious effect of off normal surface conditions in the SCC susceptibility is not yet clearly understood. A recent program was launched to get insights regarding the potential deleterious effect of off normal surface and chemistry environments on the SCC susceptibility of cold worked stainless steels. Since the presence of surface treatments due to the various steps of component manufacturing or repairing is sometimes unavoidable, a wide range of experimental techniques is deployed to reproduce some of the various off normal surface conditions. This paper summarizes the initial characterizations as well as the baseline SCC tests in PWR primary environment.

The field experience of austenitic stainless steels in PWR (Pressurized Water Reactor) primary circuits has been generally outstanding. However, the effect of so called “off normal” conditions still plays an important role regarding the SCC (Stress Corrosion Cracking) susceptibility of austenitic stainless steels (SS). Such off normal conditions can either be surface conditions or off normal water chemistry conditions, which is the topic of this paper. This paper summarizes evaluations of the SCC susceptibility of cold worked stainless steels in off normal oxygenated environment including transients. In order to study off normal chemistry conditions, SCC tests were performed in simulated PWR primary water conditions under either hydrogen water chemistry conditions or oxygenated conditions. Since the effect of transients are suspected to play a role under plant conditions, additional tests were performed with continuously changing from oxygenated to hydrogenated conditions and compared to results from tests under purely hydrogenated or oxygenated water chemistry conditions.

Sensitization and deformation have previously been implicated in the stress corrosion cracking (SCC) susceptibility of Type 304 stainless steel (SS) in oxygenated water. However, Type 304L SS, with reduced carbon content, is expected to be resistant to sensitization effects. The current work evaluates the SCC response of two dual certified Type 304/304L SSs after a sensitization heat treatment. It is shown that other material factors, namely boron content and delta ferrite stringers, can lead to sensitization and subsequent SCC even in L-grade materials.

A test program has been completed that re-examined the effect of load ratio, R, on fatigue crack growth rates (FCGRs) of austenitic stainless steel in 482 °F air and deaerated water. Data for the test program were collected at R between 0.1 and 0.95 and ΔK values between 2.5 and 50 ksi√in to ensure an overlapping dataset in R and ∆K. In contrast to the single Paris slope relationship in the ASME code, results from air tests revealed a three-regime curve across many R: a high ΔK regime similar to ASME, an intermediate ΔK regime with a decreased power law exponent and a low ΔK region where FCGR exhibit a steep downturn. Water FCGRs showed two regimes—a single power law regime and a steep low ΔK regime. FCGR sensitivity to R was greatest in the low ΔK regime for both environments.

Fatigue crack closure is widely recognized to promote fatigue crack growth retardation behavior through contacting of fracture surfaces in the crack wake and the resulting reduction in the effective stress intensity factor range (∆K) that promotes crack advance. Experimental measurements of crack closure are typically made using conventional compliance methods, but electrical potential drop has also been used to characterize crack closure behavior. Electrical potential drop measurements have detected electrical shorting across fracture surfaces of stainless steel, nickel base weld, and A508 steel tested in high temperature water environments under cyclic loads. These observations have consistently occurred under loading conditions (low R, following overloads) where closure effects are expected to be prominent and have been shown to correlate with reductions in fatigue crack growth rate. These findings suggest that electrical potential drop measurements may serve as a useful tool in assessing the influence of crack closure on corrosion fatigueCorrosion fatigue retardation behavior.

Injected vacancy, enhanced creep is hypothesized to reduce crack growth rates (CGRs) in deaerated pressurized water (DPW) in austenitic stainless steels with high sulfur levels. CGR reduction is hypothesized to occur by corrosion generated vacancy/dislocation interactions that promote dislocation climb and disrupt planar slip bands. Creep tests using tensile specimens of varying sulfur content were performed in air and DPW at 288 °C. Testing began with a hold at the flow stress, followed by fatigue cycles at room temperature (RT), then holds at flow stress and 105% flow stress. Primary creep was exhibited in the high sulfur material in DPW, after the RT fatigue cycles, and resulted in 0.19 mm of extension. Characterization revealed a corrosion product and a deformed microstructure with extensive planar slip bands in the specimen that crept. Corrosion-generated vacancies are unlikely to be the source of the primary creep. Potential mechanisms for the observed creep behavior will be discussed.

Corrosion fatigue crack growth rate tests were performed at 288 °C in high purity, deuterated water (D2O), and crack tips were examined for the presence of deuterium (D). A sample from the interior of the specimen (1T-CT) was analyzed for D and total D+hydrogen (H) using time of flight-secondary ion mass spectrometry (ToF-SIMS) and hot vacuum extraction. With SIMS, two regions 500 × 500 µm in size were analyzed. The first region was located immediately in front of the crack tip. The second region was a control, and was 6 mm away from the plastic zone. The deuterium concentration was found to be enhanced by a factor of 8.7–9.3 over the natural abundance in the plastic zone/crack tip while the concentration was enhanced by 5.9 in the bulk material. With no other source of deuterium, the detection of deuterium represents a unique marker. These data also provide evidence that hydrogen species are concentrated at the crack tip in fatigue crack growth processes at elevated temperature.

This paper describes an investigation into the mechanisms influencing environmentally assisted enhancement of fatigue crack growth of 304L stainless steel in PWR primary water. Pre-cracked specimens were tested under loading conditions containing hold periods, either using a trapezoidal waveform, or periods of saw-tooth loading interspersed with long hold periods. Several post-test characterization methods were used to provide insight into the mechanisms influencing crack growth behavior. A correlation was observed between steel sulfur content and reduced environmental enhancement, which was more pronounced under trapezoidal loading than for saw-tooth loading following extended hold periods. Post-test examination linked enhanced crack growth with highly faceted fracture surfaces, whilst lower levels of enhancement showed a less faceted and more heavily oxidized appearance. The observations suggest that, whilst enhanced corrosion due to MnS dissolution from the steel is the cause of retarded crack growth rates, different retardation mechanisms appear to contribute at high and low stress ratios. This programme was sponsored by The Electric Power Research Institute (EPRI) and a full report describing the full programme of work is available as EPRI report#3002007973.

This paper summarizes the results of a research project on environmental effects on the fatigue lifetime of selected low-alloy steels and austenitic stainless steels. Results from investigations on the effect of hold times during fatigue in air, as well as under simulated PWR primary water conditions are presented. Strain controlled fatigue tests with low strain rates and low strain amplitudes have been performed in air using a low-alloy RPV steel as well as the stabilized austenitic stainless steel 347 and the non-stabilized austenitic stainless steel 304L. Similar strain controlled tests under simulated PWR primary coolant conditions were performed using the two above mentioned austenitic stainless steels. The fatigue cycling was performed at 240 °C whereas holds were applied at higher temperature (290 °C). The effect of hold periods is compared to reference tests without hold times and to the prediction based on NUREG/CR 6909, Rev. 1.

Powder bed fusion Direct Metal Laser Melting (DMLM) is an evolving additive manufacturing (AM) fabrication technology that is providing high performance parts to many industries. This technology has significant promise for use in building components for nuclear power plants. Implementation of materials produced using this and similar processes offer a potential step change in efficiency for complex parts production and hence a potential for innovative design as well as cost savings for components in the future. Properties of AM Type 316L have been reported in previous work, showing properties that match wrought properties. The fine grain structure may even lead to better environmental resistance. However, there is a need to confirm the behavior of these innovative materials after exposure to radiation if this innovative technology is to be used in current and future nuclear applications. This paper discusses new efforts being explored via a joint program between GE Hitachi (GEH) and INL (Idaho National laboratory) aimed at developing corresponding un-irradiated and irradiated data for AM materials. This paper will present data for both Type 316L stainless steel, a single-phase alloy, and Ni-base Alloy 718, a precipitation hardened alloy, manufactured using AM. This paper, serving as a progress report, will present the mechanical property and microstructural data for both Type 316L and 718 AM alloys to assess their correspondence to wrought alloy data and establish a baseline for future comparison to irradiated properties. The paper will end by discussing the requirements for using these and other additively manufactured materials in future reactor component applications where irradiated data is not available.

Westinghouse began additively manufacturedAdditively manufactured (AM) materials research activities in 2012 to support development of advanced fuel related components. The initial objective of this work has been the fabrication and delivery of a lead test component to a Westinghouse nuclear utility customer for in-reactor insertion for a limited number of fuel cycles. It is generally recognized that a key criterion for the implementation of AM components would be a thorough understanding of the material response to neutron irradiation. Several alloys were AM fabricated, heat treated, neutron irradiated, and extensively evaluated both in the as-printed condition and in the printed and irradiated condition. Irradiation of AM miniature tensile specimens was performed at the Massachusetts Institute of Technology (MIT) Nuclear Research Reactor to 0.8 dpa at 300 °C (572 °F). Although extensive laboratory testing was performed on these materials, this paper specifically summarizes the results from room and elevated temperature tensile testing of unirradiated and irradiated AM 316L. Testing of the miniature tensile specimens was performed inside a hot cell utilizing in-cell digital image correlationDigital image correlation (DIC) and advanced video extensometryAdvanced video extensometry (AVE). Additional AM alloys are currently being irradiated in the MIT reactor to higher dpa values. The first of these samples will be shipped and subsequently tested in 2017.

Materials in nuclear power system can suffer from thermal/hydrothermal, radiation and chemical degradation due to the high-temperature, high-pressure operation condition along with the presence of water steam and radiation. One particular topic we are addressing is understanding and optimizing materials for fission gas capture. Computational modeling is an efficient tool to investigate materials behaviour in such extreme environment. Westudied a number of materials. One of these is mesoporous silica. We used a combination of Molecular Dynamics (MD) simulation and Monte Carlo (MC) simulation which were validated by detailed experiments. MD simulations reveal the porous structure transformation under high-temperature treatment up to 2885 K, suggesting the pore closure process is kinetically dependent. Based on this mechanism, we predict with the presence of water, the pore closure activation energy will be decreased due to the high reactivity between water and Si-O bond, and the materials become more susceptible to high temperature. A fundamental improvement of the material hydrothermal stability thus lies in bond strengthening. MC simulations then were used to study the the adsorption and selectivity for thermally treated MCM-41, for a variety o f gases in a large pressure range. Relative to pristine MCM-41, we observe that high temperature treated MCM 41 with its surface roughness and decreasing pore size amplifies the selectivity of gases. In particular, we find that adsorption of strongly interacting molecules can be enhanced in the low-pressure region while adsorption of weakly interacting molecules is inhibited. We have also investigated alumina as an example of a ceramic material that can be directly incorporated into the nuclear fuel itself. Unlike uranium oxide fuel, certain phases of alumina have appreciable capacity for gas absorption. The limited diffusion distance of helium and other fission product gases in the fuel may be addressed by coating micron-sized fuel particles with alumina, prior to sintering, using a unique atomic layer deposition process suitable for particles. We have investigated the feasibility of this approach using a combination of helium-focused experiments on fuel surrogate particles, together with analytical calculations of gas production rates and diffusion distances in uranium oxide. Additional studies of nanotubes of carbon and boronitride elucidated fundamental mechanisms of the influence of curvature on gas adsorption.

Martensitic stainless steels, in general, become more susceptible to Environmentally Assisted Cracking (EAC), specifically Hydrogen Assisted Cracking (HAC),Hydrogen Assisted Cracking (HAC) with increasing tensile strength (as reflected by increasing hardness). The aim of this test programme was to determine the susceptibility to HAC of a 12% chromium stainless steel as a function of material hardness, stress and environment. Incremental Step Loading (ISL) tests demonstrate a reduction in failure stress with increasing hardness due to the presence of hydrogen. Relationships between failure stress and hardness/tempering temperature are described. Testing also clearly supports the concept that there is a critical value of nominal stress, at each tempering temperature/hardness, below which HAC does not occur. Constant displacement testing results show that susceptibility to HAC is dependent upon a complex interplay between microstructure (tempering temperature/hardness), stress and environment (availability of hydrogen).

To investigate various methods to mitigate flow accelerated corrosion of carbon steels, we have deposited various metallic and composite coatings on the surface of carbon steels and tested their performance by exploiting flow accelerated corrosion (FAC) simulation experiments. From the results, we found that both Ni-P/TiO2 composite coating and Fe-based amorphous metallic coating exhibited outstanding FAC resistance thus they are expected to expand the life-time of secondary systems of nuclear power plants. Furthermore, to investigate their life-time in nuclear power plants, we investigated known mechanistic models and commercial models of FAC and imported the parameters of the coated carbon steels into the models.

The structural integrity of the reactor pressure vessel (RPV) of light water reactors (LWR) is of utmost importance regarding operation safety and lifetime. The fracture behaviour of low-alloy RPV steels with different DSA (dynamic strain aging) & EAC (environmental assisted cracking) susceptibilities and microstructures (base metal, simulated weld coarse grain heat affected zone) in simulated LWR environments was evaluated by elastic plastic fracture mechanics (EPFM) tests with different strain rates and by metallo- and fractographic post-test observations. These tests revealed some evidences of high-temperature water and hydrogen effects on the fracture behaviour and potential synergies with DSA and EAC.

Corrosion fatigueCorrosion fatigue testing of a low alloy steelLow alloy steel was undertaken to determine the effect of cycling parameters and low oxygenOxygen effects levels on the crack growth rate at various stress intensity factor ranges. Notched, pre-cracked compact tension specimens were prepared from A516-90 plate material with a sulfur content of 0.020 wt%. These specimens were tested in a water environment with a pH of ~9.0 and a temperature of 177 °C. At each stress intensity factor range, the crack growth rate was compared at three different frequencies with oxygen levels <10 to 150 ppb. At higher stress intensity factor ranges, no effect on crack growth rate from oxygen, rise time, or dwell time was observed. For the lower stress intensity factor ranges, the crack growth rate decreased with oxygen addition and additional dwell times at maximum load. The decrease in crack growth rate at lower stress intensity factors is attributed to crack tip blunting and/or crack closure effects from oxide build-up. At the higher stress intensity factors, the mechanical crack driving force was sufficient to break the oxide and continue growing the crack.

A specifically designed reference electrode was developed for analyzing the electrochemical behaviors of alloy materials in supercritical water (SCW) environments and identifying the associated electrochemical parameters. The internal Zr/ZrO2 reference electrodes constructed for high-temperature conditions were manufactured and adopted to measure the electrochemical corrosion potential (ECP) of the sample in SCW environments. Before the electrochemical analysis, the oxidation behaviour of zirconium would be investigated in SCW environment. The mass gain of zirconium is assumed due to formation of ZrO2 and there was only 78% of the original thickness of zirconium existed after the 1300 h immersion test in SCW environments with 8.3 ppm dissolved oxygen. In deaerated SCW environments, the thickness of zirconium is about 88% of the original one. The outcome indicated that the laboratory-made Zr/ZrO2 reference electrode was able to continuously operate for several months and delivered consistent and steady ECP data of the sample in SCW environments.

The influence has been studied of thermo-mechanical treatment, sensitization conditions, and neutron irradiation on the pitting corrosion resistance of austenitic 316LN stainless steel variants in 10% FeCl3·6H2O at 22 °C. Variants of this steel were modified with additions of nitrogen, manganese, copper, and tungsten, as well as testing cast, cold-rolled, grain boundary engineered (GBE), and as-received variants. It was found that the 316LN steel variant with additions of 0.2% N and 2% Mn had the best pitting corrosion resistance of all studied conditions. When irradiated in a light water reactor (LWR) to a maximum fluence of 3 × 1017 n/cm2 (E > 1.1 meV, Tirr < 50 °C), neutron irradiation surprisingly increased the resistance of GBE steels to pitting corrosion. An anisotropy of corrosion resistance of GBE and cold rolled steels was observed.

Solid particle-induced wall thinning of water-cooled nuclear power plant components is an established degradation mechanism that can affect the long term management of plant operations and component reliability. This form of material degradation is identified as erosion-corrosion and is comprised of mechanical removal known as erosion, chemical removal or corrosion, and the enhanced degradation resulting from the combined action of erosion and corrosion known as synergistic wear. Erosion-corrosion’s complicated analysis involves many factors such as fluid velocity, particle size, concentration and shape, pH, and temperature amongst several others. Erosion-corrosion is observed in heat exchanger tubing exposed to raw water (water from ponds, rivers, bays, and lakes), and steam generator blowdown piping. However, quantifying the impact of solid and liquid mixtures on light water reactor component reliability has proven to be difficult, and often underestimated. This paper describes a physics-based and probabilistic erosion model for estimating average wall thinning rates and age-dependent probabilities of exceeding user-defined wall thicknesses.

In Japan, light water reactors are built on the seacoast because they use seawater as the final heatsink. Leakage of seawater from the condenser section of the reactor could lead to contamination of the reactor coolant, and stainless steels can be susceptible to crevice corrosion in chloride-contaminated water. Therefore, it is necessary to develop counter measures for suppressing the initiation of crevice corrosion and for repassivating the corroding crevice to maintain structural reliability. To accomplish this, first the effect of molybdate anion on suppressing the initiation of crevice corrosion on 316L stainless steel in chloride-contaminated water was evaluated by potentiostatic immersion tests. Next, the effect of molybdate anion addition on the repassivation of corroding crevices was also evaluated through potentiostatic immersion tests as a function of the concentration of chloride anion. Based on the results of these examinations, the beneficial effects of the presence of molybdate anion on the suppression of initiation and propagation of crevice corrosion were quantitatively evaluated in terms of critical potentials.

Thermal desorption spectroscopy, secondary ion mass spectroscopy and scanning transmission electron microscopy have been used to investigate the effect of pH on corrosion and hydrogen pick-up behaviour in different samples of Zircaloy-4Zircaloy-4 . Samples were autoclave-oxidised in pure water and at an elevated pH (with 50% deuterated water) when compared to commercial reactors. A characteristic desorptionDesorption peak for hydrogenHydrogen has been found at ~650 °C, which occurs when the difference in free energy between hydrogen in the metal and in the gas phase becomes positive. Electron energy loss spectroscopy provided us with a method to detect and measure the thickness of the following layers (from oxide to metal): ZrO2, a previously reported ZrO suboxide, an oxygen saturated zirconium region and the Zr metal. Overall, samples exposed to a high pH show a longer time to transition and contain far less hydrogen than those oxidised in pure water. A mechanistic explanation will be provided.

The oxidation kineticsOxidation kinetics of supercritical-water-cooled reactor (SCWR) fuel cladding candidate materials, e.g. 15Cr-20Ni stainless steel (1520 SS) in supercritical waterSuper critical water reactor at 650 and 700 °C under 24 MPa has been investigated. Characteristics of oxide layers and its relation to oxidation behaviors are also studied. The applicability of the candidate materials for the fuel cladding of SCWR from oxidation kinetics, spalling susceptibility of oxide layer, and breakdown of Cr2O3 layer points of view has been discussed. The results indicate that the threshold condition for spalling of oxide layer is different at 650 and 700 °C. The decrease in oxidation kinetics of 1520 SS with time correspond to the change in rate-limiting process of oxidation from mass transfer through an Fe oxides to mass transfer through a Cr rich oxide layer with time. Based on the oxidation kinetics obtained in this study, 1520 SS is considered suitable for a fuel cladding of SCWR in combination with appropriate CW process. However, detailed evaluation and countermeasures for the degradation due to nodular oxidation are needed before application of tube-shaped 1520 SS in supercritical water at 700 °C. On the other hand, it is estimated that the use of that at 650 °C is acceptable because the weight gain after long-term exposure was considered to be much less than the threshold condition of the spalling.

The primary heat transport system of modern CANDU® (CANDU is a registered trademark of Atomic Energy of Canada Limited) reactors uses A106B piping (i.e., feeder pipes). Feeder cracking has only affected tight-radius bends at outlet feeders (higher temperature), and cracking is limited to regions with high residual stress suffering from wall-thinning by flow accelerated corrosion. To date, the mechanism of feeder cracking has not been identified. This paper includes high-resolution transmission electron microscopy and electron energy loss near edge structure characterization of inside and outside surface intergranular cracks from ex-service CANDU feeders. Prior to this work, no high resolution characterization has been performed for CANDU feeder cracking. All intergranular cracks show evidence of cementite decomposition, leading to decoration of grain boundaries with amorphous carbon, and carbon diffusion along un-cracked boundaries ahead of crack tips. Sulfur has been found on the oxide-metal interface of all intergranular cracks, but is not observed ahead of the crack tips. Sulfur is believed to be from the breakdown of manganese sulfides during service. The cementite decomposition and breakdown of manganese sulfides are believed to be accelerated in the presence of hydrogen produced from the flow accelerated corrosion. Small (<15 nm) voids are also present ahead of some intergranular crack-tips along the ferrite-ferrite boundaries, indicating that hydrogen enhanced, low temperature creep-cracking, may also contribute to intergranular fracture.

Elevated temperature is the primary source of aging for nuclear power plant electrical cable insulation and jacketing, but gamma radiation is also a significant contributor to structural changes that result in loss of polymerPolymers mechanical and electrical properties in affected plant locations. Despite many years of research, the combined degradation effects of simultaneous exposure to thermal and radiation stresses are not well understood. As nuclearNuclear operators prepare for extended operation beyond initial license periods, a predictive understanding of exposure-based cable degradation is becoming an increasingly important input to safety, licensing, operations and economic decisions. We focus on carefully-controlled simultaneous thermal and gamma radiation aging and characterization of the most common nuclear cable polymers to understand relative contributions of temperature, time, dose and dose rate to changes in cable polymer material structure and properties. Improved understanding of cable performance in long term operation will help support continued sustainable nuclear power generation.

This paper describes the use of Principal Component Analysis (PCA)Principal Component Analysis (PCA) as a statistical method to identify key indicators of degradation in nuclear power plant cable insulation. Seven kinds of single-point data and four kinds of spectral data were measured on cross-linked polyethylene (XLPE) that had undergone aging at various doses and dose rates of gamma radiation from a cobalt 60 source, and various elevated temperatures. To find the key indicators of degradation of aged cable insulation, PCA was used to reduce the dimensionality of the data set while retaining the variation present in the original data set. For example, PCA reveals that, for material aged at 90 °C, elastic modulus shows a positive correlation with total dose while mass loss, oxidation induction time and density show negative correlations with the same parameter.

Low voltage (LV) power,Low voltage cables control and instrumentation cables are essential to the safe and reliable operations of nuclear power plants (NPPs). However, considering the huge consumption of cables in NPPs, it is impractical and cost prohibitive to replace cables when they reach the end of their design life. As a result, condition monitoringCondition monitoring and agingAging management of cables is critically important for the life extension of NPPs. The aging of LV cables is characterized by the degradation of the polymeric insulation and jacket materials leading to their mechanical failure or the loss of their ability to withstand critical conditions. This paper focuses on studying material characterization techniques (already existing and new) to monitor the change inPolymer characterization polymeric material propertiesMaterial properties and also to establish a general framework between material properties and ageing conditions that could assist in predicting the condition of cables and estimating their remaining life.

For medium voltage (MV) Medium voltage cables power cables with voltages between 5 and 35 kV that provide power to emergency and safety support systems in nuclear power plants (NPPs), degradation and failure of cables due to water exposure has occurred. Systematic studies of current NPP MV cable systems are being carried out to determine and characterize cable-aging mechanisms and provide greater accuracy to models being developed to predict MV cable field performance to support existing cable aging management and monitoring programs. To validate models based on representative laboratory specimens, samples from harvested MV cable systems are being subjected to accelerated aging and degradation from humid conditions and submergence. Degradation will be characterized via partial discharge and voltage endurance testing of the cable and induced water tree growth in insulation. The technical approach to be used for the testing harvested MV cable samples is presented.

Cable insulation polymers are among the more susceptible materials to age-related degradation within a nuclear power plant. This is recognized by both regulators and utilities, so all plants have developed cable aging management programs to detect damage before critical component failure in compliance with regulatory guidelines. Although a wide range of tools is available to evaluate cables and cable systems, cable aging management programs vary in how condition monitoring and nondestructive examination is conducted as utilities search for the most reliable and cost-effective ways to assess cable system condition. Frequency domain reflectometry (FDR)Frequency domain reflectometry is emerging as one valuable tool to locate and assess damaged portions of a cable system with minimal cost and only requires access in most cases to one of the cable terminal ends. This work examines a physics-based model of a cable system and relates it to FDR measurements for a better understanding of specific damage influences on defect detectability.

Aging mechanismsAging mechanisms and a nondestructive aging indicator of filled cross-linked polyethylene (XLPE)Filled cross-linked polyethylene cable insulation material used in nuclear power plants (NPPs) are studied. Using various material characterization techniques, likely candidates and functions for the main additives in a commercial filled-XLPE insulation material have been identified. These include a mixture of brominated components such as decabromodiphenyl ether and Sb2O3 as flame retardants, ZnS as white pigment and polymerized 1,2-dihydro-2,2,4-trimethylquinoline as antioxidant. Gas chromatography-mass spectrometryGas chromatography-mass spectrometry , differential scanning calorimetryDifferential scanning calorimetry , oxidation induction timeOxidation induction time and measurements of dielectric loss tangentDielectric loss tangent are utilized to monitor property changes as a function of thermal and radiation exposure of the cable material. The level of antioxidant decreases with aging by volatilization and chemical reaction with free radicals. Thermal aging at 90 ℃ for 25 days or less causes no observable change to the cross-linked polymer structure. Gamma radiation causes damage to crystalline polymer regions and introduces defects. Dielectric loss tangent is shown to be an effective and reliable nondestructive indicator of the aging severity of the filled-XLPE insulation material.

We have succeeded in detecting the degradation of cable’sCable polymeric insulationPolymeric insulation well before its continual use becomes risky. Degradation of organic polymers is mainly caused by oxidation if the ambience around the cable contains oxygen. When organic polymers are oxidized, polar carbonyl groups are formed, by which the permittivity is increased. This in turn decreases the characteristic impedanceCharacteristic impedance of a polymer-insulated cable. If we inject electromagnetic waves in a very wide frequency range into the cable and measure the ratio of reflected power to injected power, the information on the effects of the characteristic impedance changes is included in the frequency spectra of the ratio. If we do inverse Fourier transform, we can convert the data to a time domain. Therefore, we can know the degraded portion by multiplying the velocity of electromagnetic waves in the cable.

Cross-linked polyethylene (XLPE) is used widely as insulation material in low-voltage instrumentation cables deployed in nuclear power plants (NPPs). Suffering from degradation due to exposure to heat and radiation, the insulating properties of the material gradually degrade, which may imperil the safe operation of NPPs. In this paper, a new capacitive sensing method is introduced for nondestructive evaluation of aged instrumentation cable with XLPE insulation. The capacitive sensor is capable of measuring capacitance of cable materials and may potentially be employed to extract the dielectric properties of insulation material from those of the entire cable. Interdigital capacitive sensor designs are targeted that maximize sensitivity to changes in the dielectric values of the cable polymers.

Cross-linked polyethylene (XLPE)XLPE cable insulation samples were exposed to heat and gamma radiation at a series of temperatures, dose rates, and exposure times to evaluate the effects of these variables on material degradation. The samples were tested using the solvent incubation method to collect gel fractionGel fraction and uptake factorUptake factor data in order to assess the crosslinkingCrosslinking and chain scission occurring in polymer samples with aging. Consistent with previous reports, gel fraction values were observed to increase and uptake factor values to decrease with radiation and thermal exposure. The trends seen were also more prominent as exposure time increased, suggesting this to be a viable method of tracking structural changes in the XLPE-insulated cable material over extended periods. For the conditions explored, the cable insulation material evaluated did not indicate signs of anomalous aging such as inverse temperature effectInverse temperature effect in which radiation-induced aging is more severe at lower temperature. Ongoing aging under identical radiation conditions and at lower temperature will further inform conclusions regarding the importance of inverse temperature effects for this material under these conditions.

Silicone rubber (SiR)SiR was gamma irradiated at 125, 145 and 185 °C or thermally aged at 220, 250 and 280 °C and the resultant changes in performance were evaluated. It has become clear from instrumental analyses that crosslinking via oxidation of silicon atoms and chain scission are induced by gamma raysGamma rays . Furthermore, from the temperature dependence of real relative permittivity at high frequencies, the thermal expansion coefficient was found to become smaller with the increase in dose. These results can be understood well by the chemical and structural changes in SiR induced by the degradationDegradation .

This paper documents the development of signal processing and machine learning techniques for the detection of Alkali-silica reaction (ASR). ASR is a chemical reaction in either concrete or mortar between hydroxyl ions of the alkalis from hydraulic cement, and certain siliceous minerals present in some aggregates. The reaction product, an alkali-silica gel, is hygroscopic having a tendency to absorb water and swell, which under certain circumstances, leads to abnormal expansion and cracking of the concrete. This phenomenon affects the durability and performance of concrete cause significant loss of mechanical properties. Developing reliable methods and tools that can evaluate the degree of the ASR damage in existing structures, so that informed decisions can be made toward mitigating ASR progression and damage, is important to the long-term operation of nuclear power plants especially if licenses are extended beyond 60 years. The paper examines the differences in the time-domain and frequency-domain signals of healthy and ASR-damaged specimens. More precisely, we explore the use of the Fast Fourier Transform to observe unique features of ASR damaged specimens and an automated method based on Neural Networks to determine the extent of ASR damage in laboratory concrete specimens.

The Grizzly code is being developed under the US Department of Energy’s Light Water Sustainability program as a tool to model aging mechanisms and their effects on the integrity of critical nuclear power plant components. An important application for Grizzly is the modeling of aging in concrete structures, which can be due to a number of mechanisms. Initial focus in this area has been on modeling expansive reactions due to alkali-silica reactions or radiation-induced volumetric expansion. Grizzly is an inherently multiphysics modeling platform that naturally permits including the effects of multiple coupled physics in a simulation. Models have been developed for transport of heat and moisture in concrete, and these have been coupled and used as inputs to models for expansive reactions. This paper summarizes this capability, and demonstrates it on a representative structure.

The Electric Power Research Institute (EPRI) has been engaged in collaborative research and development activities related to concrete in nuclear applications over the past several years in concert with the nuclear generation industry, foreign and domestic national laboratories and regulatory bodies and universities. The EPRI Long Term Operations program is focused on performing research activities that will help the industry extend operation beyond the first period of license renewal, which for US plants means operation beyond 60 years. In this overview talk, three subjects will be addressed—radiation damage in boiling and pressurized water reactor concrete biological shields, boric acid attack of pressurized water reactor spent fuel pool concrete substructures and alkali-silica reaction degradation of concrete structures. The results of these and other studies are expected to support utilities as they demonstrate technical bases to regulatory bodies for long term operation of commercial nuclear plants.

Plans for extended operation of US nuclear power plants (NPPs) beyond 60 years have resulted in a renewed focus on the long-term aging of materials in NPPs, and specifically on reactor cavity concrete. To better understand the effects of neutron irradiationNeutron irradiation on reactor cavity concrete, a select group of mineral analoguesMineral analogues of concrete aggregatesAggregates were irradiated at the Oak Ridge National Laboratory High Flux Isotope Reactor at three different fluence levels and at two temperatures. The purpose was to investigate the degradation of mechanical properties at neutron doses above the levels expected in US NPPs under extended operation. Preliminary findings using nanoindentation clearly show that changes in the mechanical properties of these minerals can be observed and correlated to the neutron-induced damage. Scanning electron microscopy reveals changes in deformation and fracture mechanisms in the irradiated mineral analogies. Results for the nanohardnessNanohardness as a function of dose and temperature are presented and discussed for quartz, calcite, and dolomite.

FeCrAl alloys are rapidly becoming mature candidate alloys for accident tolerant fuel applications. The FeCrAl material class has shown excellent oxidation resistance in high-temperature steam environments, a key aspect of any accident tolerant cladding concept, while also being corrosion resistant, stress corrosion cracking (SCC) resistant, irradiation-induced swelling resistant, weldable, and formable. Current research efforts are focused on design, development and commercial scaling of advanced FeCrAl alloys including large-scale, thin-walled seamless tube production followed by a broad spectrum of degradation evaluations in both normal and off-normal conditions. Included in this discussion is the theoretical analysis of the alloying principles and rules, alloy composition design, and overview of the most recent empirical database on possible degradation phenomena for FeCrAl alloys. The results are derived from extensive in-pile and out-of-pile experiments and form the basis for near-term deployment of a lead-test rod and/or assembly within a commercially operating nuclear power plant.

Advanced steels, including FeCrAl are being considered as an alternative to the standard light water fuel (LWR) cladding, Zircalloy. FeCrAl has superior mechanical and thermal properties and oxidation resistance relative to the Zircalloy standard . Uranium Silicide (U3Si2) is a candidate to replace uranium oxide (UO2) as LWR fuel because of its higher thermal conductivity and higher fissile density relative to the current standard, UO2. The interdiffusion behavior between FeCrAl and U3Si2 is investigated in this study. Commercially available FeCrAl, along with pellets fabricated at the Idaho National Laboratory were placed in diffusion couples. Individual tests have been run at temperatures ranging from 500 ℃ to 1000 ℃ for 30 h and 100 h. The interdiffusion is analyzed with an optical microscope and scanning electron microscope (SEM). Uniform and planar diffusion regions along the material interface are illustrated with backscatter electron micrographs and energy-dispersive X-ray spectroscopy (EDS).

The US Department of Energy is working with commercial fuel vendors to develop advanced technology or accident tolerant fuels (ATF) for the current fleet of light water power reactors. General Electric and Oak Ridge National Laboratory are evaluating the concept of using iron-chrome-aluminum (FeCrAl) alloys as cladding for the current fuel of uranium dioxide pellets. In the case of a loss of coolant accident, the reactor may need to be flooded with fresh water when the cladding could be in the temperature range above 1000 ℃. It is important to determine the integrity of the cladding material after being quenched in water. Tests were performed for six alloys of interest which were exposed for 2 h at 1200 ℃ in air, argon or steam and then quenched in ambient temperature water. The resulting mechanical properties were evaluated and compared with the mechanical properties of the as received material. The FeCrAl alloy retains its yield strength after the high temperature excursions, with minimal oxidation but with some loss of ductility.

FeCrAl alloys are promising for developing accident tolerant nuclear fuel claddings. These alloys showed good environmental compatibility and oxidation resistance in elevated-temperature water and steam, as well as low radiation-induced swelling. However, FeCrAl alloys may suffer from several degradation mechanisms, one of which may be a susceptibility to cracking during welding. Here, a set of advanced modified FeCrAl alloys were designed and produced by varying Al-content and employing additions of Nb and TiC. Strength, ductility, and deformation hardening behavior of the advanced FeCrAl alloys and their weldments are discussed.

Thermal energy release and hydrogen generation due to breakaway oxidation of Zr fuel cladding materials are of concern in accident scenarios involving extreme temperature increase (up to 1200 °C). As a result, potential accident tolerant fuel cladding (ATFC) materials and coatings are being investigated. Physical vapor deposited CrN coatings are considered as possible protective barrier materials for Zircaloys. In addition, Fe–Cr-Al alloys are considered potential candidate materials for ATFC due the formation of protective alumina at high temperatures which maintains resistance by preventing oxide breakdown. Both CrN-coated Zircaloys and a Fe–Cr-Al model alloy were exposed to 300 °C water and steam environments up to 1200 °C to evaluate their resistance to corrosion under normal reactor operating conditions and to high temperature steam oxidation. Surface analytical techniques are used to evaluate the effectiveness of oxides and/or coatings over the 300 °C water to 1000 °C steam temperature regime.

In order to better understand the high temperature steam oxidation behavior ofFeCrAl FeCrAl alloys, this study addressed two topics. The first is continuing to evaluate the effect of alloy composition on performance of commercial and laboratory-made candidate FeCrAl alloys. For a few optimized compositions, this includes the performance of commercially-made tubing where it is clear that dropping the Cr content from 20% to 10–13% reduces the maximum operating temperature in steamSteam by ~50 °C. The second addresses more realistic accident conditions. Model FeCrAl compositions that were exposed in ~300 °C water for 1 year were subsequently “ramp” tested in steam at 5 °C/min to 1500 °C to assess the effect of the Fe-rich oxide formed in water on the subsequent steam oxidation resistance. For Fe-18Cr-3Al+Y, the 1 year exposures in three different LWR water chemistries did not affect the ability to form alumina to 1500 °C. However, for marginal alloys Fe-13Cr-4Al and Fe-10Cr-5Al, some specimens began forming voluminous Fe-rich oxide at lower temperatures.

Irradiation enhanced corrosionCorrosion behavior of Accident Tolerant Fuel candidate alloys T91 and Fe15Cr4Al were evaluated using in-situ proton irradiation-corrosion experiments in hydrogenated pure water (at 320 °C, 3 wppm H2) with a 5.4 MeV protonProton beam. The thin sample acted as a “window” to allow protons to fully penetrate the sample while maintaining system pressure. The area of the samples exposed to the proton beam experienced effects from displacement damage and radiolysis products. The aim of the study was to characterize the effect of radiation on the kinetics and character of oxidation caused by accelerated waterside corrosion under irradiationIrradiation . Samples irradiated with protons for total displacement damage of ~0.1 dpa (dose rate in water, 400 kGy/s) at an exposure time of 24 h were compared. Oxide morphology, phase structure, and composition of the oxide, metal and the metal/oxide interface were investigated using TEM and EDS and are related to the test conditions. The oxidation rate or resulting oxide thickness is dependent on the alloy Cr content; the oxidation rate increased as the Cr content decreased. The resulting oxide consists of an inner layer of Cr-rich spinel oxide and outer magnetite crystals in the unirradiated region; while the irradiated region consists of Cr-rich inner oxide spinel that was partially dissolved and coverage of outer non-faceted hematite precipitates.

As a candidate material for accident-tolerant fuelAccident-tolerant fuel cladding for light water reactors (LWR), SiCf–SiC composite materials possess many attractive properties. However, prior work has shown that SiC is susceptible to aqueous dissolution in LWR coolant environments. To address this issue, candidate coatings have been developed to inhibit dissolution. For this study, CVD SiC samples were prepared with Cr, CrN, TiN, ZrN, NiCr, and Ni coatings. Uncoated SiC and SiCf–SiC samples were also prepared. The samples were exposed for 400 h in 288 ℃ water with 2 wppm DO in a constantly-refreshing autoclave to simulate BWR–NWC. Cr and Ni coated samples lost less mass than the uncoated SiC sample, indicating an improvement in performance. The CrN coating resisted oxidation, but some of the coating was lost due to poor adhesion. The TiN coated sample gained significant mass due to oxidation of the coating. ZrN and NiCr coatings showed significant corrosion attack. SiCf–SiC ceramic matrix composite materials dissolved much faster than the CVD SiC sample, demonstrating the need for mitigation coatings if CMCs are to be used in LWRs. This work demonstrates the promise of Cr, Ni and CrN coatings for corrosion mitigation in LWRs, and shows that NiCr and ZrN are not promising coating materials.

Accident-tolerant fuel (ATF)Accident-Tolerant Fuel (ATF) is an increasingly important research topic for the nuclear industry, and ceramics such as SiCSiC are strong contenders for deployment as ATF cladding. The hydrothermal corrosion Hydrothermal corrosion characteristics of SiC and Al2O3 Al2O3 were investigated via constantly-refreshing autoclave corrosion and post exposure characterization. Four different types of chemical vapor deposited (CVD) SiC specimens were examined (two with high electrical resistance, one with low electrical resistance, and a single crystal 4H structural variant). Al2O3 specimens were prepared in single crystal and polycrystalline states. PWR primary water, BWR–HWC, and BWR–NWC environments were maintained throughout the experiments. Characterization conducted using SEM and EDS was used to determine factors affecting corrosion rates and susceptibility to grain boundary attack in each water chemistry condition. Raman spectroscopy was also used to determine chemical variation of the surface with corrosion. Grain boundary attack was found to be significant for both alumina and SiC polycrystalline variants.

The corrosion behavior of ceramic coated ZIRLO tubing was evaluated in supercritical water to determine its behavior in high temperature water. The coating architecture consisted of a 4 bilayer TiAlN/TiN coating with Ti bond coat on Zirlo tubes using cathodic arc physical vapor deposition (CA-PVD) technique. On exposure to deaerated supercritical water at 542 °C for 48 h coated tubes exhibited significantly higher weight gain compared to uncoated Zirlo. Examination revealed formation of a uniform ZrO2 layer beneath the coating and of a thickness similar to that on the uncoated tube inner surface. The defects generated during the coating process acted as preferential paths for diffusion of oxygen resulting in the oxidation of substrate Zirlo. However, there was no delamination of the coating.

Many Stress Corrosion Cracking (SCC) models have been developed so far. Quantitative empirical models, trying to predict initiation and crack growth rate of nickel alloys exposed to pressurized water reactor primary water do not describe physical mechanism and suffer a lack of accuracy. By contrast, models describing the possible involved physical mechanisms responsible for degradation (selective oxidation of grain boundaries in the case of Alloy 600Alloy 600 exposed to PWR primary water) are usually qualitative. In the current paper, a ‘local’ model is proposed to better predict SCC. In order to succeed, behaviors assumed to be involved in the SCC process were calibrated and coupled: intergranular oxidation rate, intergranular stresses, resistance to cracking of oxidized grain boundaries. The output of the model is the time to reach a given crack depth. This paper introduces the first calibration of parameters for Alloy 600 exposed to primary water.

A numerical approach was developed to predict Stress Corrosion Cracking (SCC) kinetics and location, in 3D. The calculation is a fast post finite element modeling analysis chaining IGSCC initiation and crack growth models. Firstly, the proposed methodology offers the possibility to optimize the calibration of models, when SCC tests are simulated. Secondly, calibrated models can be used to predict SCC in large structures. In the proposed paper, benefits and limitations of the methodology will be introduced.

We investigated the applicability of a stress corrosion cracking (SCC) engineering model and simulation method developed on the basis of the SCC of sensitized 304 stainless steel in a simulated BWR environment to the primary water stress corrosion cracking (PWSCC). We conducted a uniaxial constant loading test on Alloy 600 in a 400 °C hydrogenated steam environment and found that the number of cracks observed on a specimen surface after every passage of 450 h could be approximated to Poisson distribution, indicating that a Poisson random process model is applicable to the SCC in this system. By applying the engineering model, we statistically processed experimental data by assuming that the time distribution of occurrence of microcracks follows exponential distribution, and then obtained input data for the SCC simulation. Using coalescence coefficient, k, as a fitting parameter to obtain a reasonable k-value, it was found that the best agreement between the experimental and simulation results for the number of microcracks and the maximum crack length at k = 0.15. This is about one third the k-value of 0.5 found in sensitized 304 stainless steel in the BWR environment, indicating that coalescence is more subdued in PWSCC than in SCC in the BWR environment.

Austenitic 304 stainless steels (SS) and carbon steels (CS) are widely used as structural materials for components and pipe assemblies in nuclear power plants. Steels also act as the important canister materials in the dry storage of spent nuclear fuels. However, it is well known that stainless steels and carbon steels are susceptible to stress corrosion cracking (SCC) in certain environments induced by sea salt particles and chlorides. It is known that the TiO2 coating acts as a non-sacrificial anode and protects the steel cathodically under UV illumination. In this study, the photoelectrochemical behavior of the steel with TiO2 coating by sol-gel method was investigated to mitigate atmospheric SCC. In addition, the slow decline of photovoltage of the samples coated with TiO2 after UV illumination have been studied. These results indicate that the TiO2 treatment with or after UV illumination would effectively reduce the steel corrosion rate in atmospheric environments.

Baffle-former boltBaffle-former bolt failures have been observed in recent inspections of pressurized water reactors (PWRs). These failures are understood to be primarily caused by irradiation-assisted stress corrosion cracking (IASCC). A prognostic method has been developed for simulating the degradation of baffle-former bolts due to IASCC. The method characterizes the evolution of stress in a reactor environment, as well as the redistribution of stress amidst neighboring bolt failures. Empirically-validated Weibull parameters are utilized in a stochasticStochastic framework, and the model is exercised as a Monte Carlo simulation to evaluate a range of plausible scenarios from which trends of bolt failure rates and patterns can be determined. This semi-empiricalSemi-empirical methodology is informed by both operating experience and a more detailed, predictive finite element analysis model that encompasses the full range of phenomenological effects common to the operating environment within a PWR.

Alloy 82 weld metal shows higher Stress Corrosion Cracking (SCC) resistance in BWR environments than Alloys 182 and 132. To define the relative factor of improvement in SCC resistance and its impact on plant management properly, it is appropriate to establish a SCC growth rate disposition curve for Alloy 82 in BWR environments. In this study, several factors that influence SCC growth behavior of Alloy 82 are evaluated based on the latest Crack Growth Rate (CGR) data collected in Japan BWR Owners Group projects. The goal is to provide a technical basis on which the validity of the data will be evaluated prior to proposing a new disposition curve. The factors evaluated include effects of type of Alloy 82 weld, specimen size, post weld heat treatment (PWHT) and sulfate addition on SCC growth rate.

The effect of stress intensity factor, cold workCold work , corrosion potential and water purity on the stress corrosion crack (SCC) growth rate behavior of as-received and as-received plus 19.3% cold worked XM-19 was investigated in 288 °C BWR water. For 19.3% cold rolled XM-19, high to very high crack growth ratesCrack growth rate were consistently observed at high corrosion potential, largely independent of heat or orientation. As received XM-19 exhibited SCC growth rates ~5–10X slower than cold worked XM-19, but these rates are still considered high. For all materials and conditions, low corrosion potential conditions reduced the growth rates by about an order of magnitude, and somewhat more if impurities were present in the water. The SCC growth rates for both conditions of XM-19 were somewhat higher than the equivalent conditions of 18-8 stainless steels, such as Types 304/304L and 316/316L. Higher growth rates tend to be observed at higher yield strength, and XM-19XM-19 stainless steel has an elevated yield strength from nitrogen-strengthening; incomplete annealing in the as-received material can also increase the yield strength. The J-R fracture resistanceJ-R fracture resistance of 19.3% cold rolled XM-19 and as-received XM-19 in multiple orientations and with replicates was evaluated in 288 °C air. The data show a significant effect of crack orientation in the plate (the rolling plane coincides with the plane of the plate), consistent with the inhomogeneous nature of the microstructure. The fracture resistance of as-received XM-19 was good, but the 19.3% cold rolled XM-19 specimens exhibited low toughness, to the extent that many tests were invalid. Fracture resistance in 80–288 °C water environments was not evaluated, but is relevant to LWR components. Irradiation of this heat of XM-19 is in progress at the Idaho National Laboratory Advanced Test Reactor.

Nickel AlloyNickel based alloys X-750X-750 is a Ni-Cr–Fe alloy with good corrosionCorrosion properties and high strength at elevated temperature. It is commonly used for spacer grids in Boiling Water Reactors (BWR)BWR . In this environment, the material can suffer from significant corrosion, leading to weight loss by metal dissolution. To further improve the characteristics of this material, a process called preoxidation is often performed. This results in the formation of strengthening γ’-Ni3(Ti, Al) precipitates and a thin oxideOxide formation on the surface. In this paper, preoxidized and non-preoxidized specimens are compared with respect to their oxidation properties. We report about microstructural studies made on specimens exposed in simulated BWR environment for 24 h and 840 h. Electron microscopy techniques have been used to investigate the oxide microstructures. A comparison between these specimens shows the complexity of the corrosion process and the impact of preoxidation. Preoxidized specimens show thinner and more homogenous oxides than non-preoxidized ones. They lose less mass and build thinner oxides. The preoxidation layer consists of a bilayer oxide of NiFe2O4 NiFe2O4 and Cr2O3 that is preserved during the long exposure. NiFe2O4 spinel crystals are present on the surface of all exposed specimens, a result of re-precipitation of dissolved metal ions.

This paper contributes to a TEM examination on the oxide films formed at three locations along a crack path in Alloy 182 weld from a BWR core shroud support leg, namely, the crack mouth, the midway between the mouth and the crack tip, and the crack tip. In the crack mouth the oxide film was approximately 1.6 μm in thickness and consisted of relatively pure NiO. The midway oxide film was mainly a nickel chromium oxide with a film thickness of 0.3 μm. At the crack tip the oxide film was a nickel chromium iron oxide with a film thickness of 30 nm. In all studied locations the main oxides had the similar rocksalt structure and the cracks were much wider than the thicknesses of the oxide films. It probably suggests that the corroded metal was largely dissolved into the coolant. The different dissolution rates of nickel, chromium and iron cations in the oxide films are clearly displayed with the compositions of the residual oxides. The oxide stability under different redox potentials along the crack path is briefly discussed.

The objective of this study was to quantify the effect of chloride transients on stress corrosion cracking of pressure vessel low alloy steels. Two heats of reactor pressure vessel steel were evaluated at various chloride concentrations in both NWC and HWC environments. The tests showed that low alloy steels can exhibit a delayed cracking response during a chloride transient. The delayed response is attributed to the concentrating and dilution processes of anionic impurities inside the crack. The crack can maintain its original growth rate for a certain period after the chemistry change. The effects of chloride concentration, stress intensity factor (K) and periodic load cycling on the crack incubation and growth in low alloy steel will be discussed. The results from this work will provide a direct input to development of a crack growth model for low alloy steels, water chemistry guidelines and effects of chloride transients on crack growth.

As reactor startup begins, the ECP is initially high in the oxygenated water environment established during a cold shutdown. Consequently, the components would exhibit the higher crack initiation and propagation rates of IGSCC during startup period than in the remainder of the cycle. The corrosion current density response of stainless steel exposed to H2O2 was larger than that of those exposed to O2, and it remained at a higher value even at the low level of several ppb. As noble metal was applied in the BWRs to catalyze the chemical reactions of H2O2 and O2, this study evaluated the corrosion behaviors of both oxidants on the components of stainless steel. The corrosion potentials and corrosion current densities of 304SS with Pt coating were investigated in pure water with dissolved oxygen or hydrogen peroxide concentrations at various temperatures.

Zinc injection in reactor feed water is a well-known mitigation strategy for prevention of radioactive 60Co deposition in both Boiling and Pressurised water reactors. Furthermore, zinc leads to the formation of a thinner, more stable oxides arising from the thermodynamically driven replacement of Ni and Fe in the characteristic spinel type oxide formed on stainless steel. However, the interaction of zinc with the oxide formation under different water chemistries is not fully understood. Oxidation tests on type 316 stainless steel were performed under two hydrogen water chemistry conditions (HWC), with and without zinc injection and the resultant oxides analysed using analytical electron microscopy (AEM), field emission gun scanning electron microscopy (FEG-SEM), and energy dispersive X-ray spectroscopy (EDXS). The present work identifies and quantifies the positive microstructural changes that Zn has on the oxide formation on a #600 grit surface and an OPS polished 316 SS surface under boiling water reactor (BWR) conditions.

Noble metalNoble metal injection is widely used to mitigate stress corrosion cracking (SCC) of reactor components. Despite its wide use, there are still open questions regarding the parameters affecting the application process and possible improvements to it. Laboratory experiments in a high-temperature water loop at PSI were complemented by exposure of specimens in the mitigation monitoring system (MMS) at KKL. The influence of parameters such as flow conditions, structural material composition, surface roughness and geometry on the deposition behavior of the platinum (Pt)Platinum nanoparticlesNanoparticles was investigated. Furthermore, the long-term stability of the coverage of surfaces by Pt particles was analyzed. The composition of the underlying alloy was found to have an effect on the deposition behavior, whereas surface roughness has no measurable impact. Pt showed a limited durability on steel surfaces and, after the end of the application, the remobilized Pt seems to re-deposit only minimally on nearby surfaces.

The development and implementation of On-Line NobleChemTM (OLNC) has now occurred for many years with almost the entire US BWR fleet, including Mexico, and some European BWRs using this process to achieve mitigation of IGSCC. The data confirming the effectiveness of OLNC has indirectly shown significant reductions in incidences of crack initiation as well as the suppression of growth of any previously existing cracks. However, the robustness of the process and its ability to protect all core and lower plenum regions requires on-going confirmation of the presence of nanometer sized platinum particles as well as verifying the adequacy of the coverage on surfaces of the core structural materials. Efforts by GE Hitachi Nuclear Energy (GEH) have continued in cooperation with the BWR utilities to both properly inject the platinum solution into the coolant as well as to monitor the surface characteristics of actual core components. This paper will show the results of on-going efforts to confirm platinum coverage of stainless steel surfaces using Field Emission Scanning Electron Microscopy (FE-SEM) methods. Additional efforts to tie all the process methods together to establish and confirm OLNC performance will also be presented. Finally, the paper will review continuing plans moving forward to validate OLNC performance across the BWR fleet.

It is necessary to study crevice structures which can mitigate crevice corrosionCrevice corrosion as the origin of SCC of materials used in BWRBWR core region. A fundamental crevice corrosion simulation model has been developed to design corrosion control structures for these materials. Effects of the width and the depth on the corrosive environment in a crevice were studied based on that model. Calculated pH in the crevice decreased with time for all crevice geometries. The lowest pH was found at the deepest position in the crevice for all the cases. It seemed there was only a negligible difference in pH where the crevice depth was deeper than the specific depth which depended on the crevice width.

In situ electrochemical impedance spectroscopy measurement within crevice of stainless steel in 288 °C water has been conducted to analyze crevice water chemistry. Small sensors ( $$ {\varphi} {\sim} 250\,\upmu{\text{m}}) $$ measured local solution electrical conductivity κ crev, polarization resistance and electrochemical corrosion potential. Real-time response of the κ crev as functions of bulk water conductivity and dissolved oxygen (DO) concentration has been quantitatively analyzed. The κ crev differ more than an order of magnitude depending on the oxygen potential inside the crevice. The κ crev increased with addition of small amount of bulk DO (e.g. 30 ppb). The maximum κ crev was observed with DO of 32,000 ppb and became more than 100 times higher than that of bulk water. The effect of geometrical factors on the crevice environment was also found to play an important role in the water chemistry inside.

In-service inspections of Zr-2.5NbZr-2.5Nb pressure tubes may reveal blunt flaws such as fretting wear or crevice corrosion marks. These flaws pose no immediate threat to the integrity of the pressure tube but may be potential fatigue crack initiationCrack initiation sites. An understanding of the effect of the coolant environment, specifically on fatigueFatigue crack initiation, is important in this context. Tests were conducted on notched transverse tensile specimens at 275 and 300 °C with a load rise time between 50 and 3600 s. Current tests investigated the effects of applied loading frequency and hydrogenHydrogen on fatigue crack initiation. Results have indicated that long rise time and a water environment reduce the time to fatigue crack initiation in non-hydrided and pre-hydrided specimens as compared to tests conducted in air. If enough hydrogen is able to diffuse to the notch during the test, it may also be possible to reach conditions where there is an interaction between corrosionCorrosion , fatigue and hydride cracking.

Hydride precipitation in zirconium claddingZirconium cladding is known to cause severe loss of toughness and greatly increase the risk of mechanical failure and fuel leakage. Modeling hydride formation kinetics is critical to the safety assessment of the fuel-cladding system and the entire reactor system. Existing reduced order models do not provide such details as number density and size distribution of hydride precipitates. We have recently developed a cross-scale cluster dynamics modelCluster dynamics modeling with increased physical details and enhanced predictive capability for the hydride formation kinetics in zirconium. Our model takes information from atomistic simulations, such as migration energy of interstitial hydrogenHydrogen and formation/binding energy of hydride embryos/clusters, as input, and establishes and solves a system of rate equations that describe the evolution of concentrations of freely migrating hydrogen as well as sessile hydride clusters of all different sizes. Used here to simulate an in situ hydride growth experiment on a TEM, our model is able to reproduce the linear growth behavior of pre-existing hydrides under hydrogen ion implantation and provide possible explanations for the estimated growth rate.

Correctly predicting the mechanical behavior of zirconium fuel cladding during a LOCA transient is critical for nuclear safety analysis as the fuel rod needs to maintain its coolable geometry throughout the LOCALOCA sequence. A physically-based zirconium alloy corrosion model called the Coupled Current Charge Compensation (C4) is developed. The model calculates the coupling of oxygen, electron and hydrogen currents and predicts the oxide, oxygen-stabilized $$ \alpha $$ -Zr and prior- $$ \beta $$ -Zr layers kinetics as well as the oxygen concentration profiles during a LOCA scenario. The results obtained during isothermal conditions are compared to experimental data for validation. Future developments of the C4 model include an implementation into the nuclear performance code BISON, which currently does not provide a physical description of the oxygen and hydrogen concentration profiles in the cladding. Thanks to the C4 implementation into BISON, structural integrity of the fuel cladding following a LOCA event can be assessed.

The corrosion and hydrogen pickup of in-reactor zirconium-based cladding is currently limiting the maximum fuel burnup in light-water reactors. Since the oxidation rate and hydrogen pickup fraction of zirconium alloys vary significantly as a function of exposure time, burnup, and alloy composition, it is critical to better understand the underlying mechanisms to model and predict corrosion behavior. Following the analysis of ~500 autoclave coupons, a physically based zirconium-alloy corrosion model founded on first principles, named “Coupled Current Charge Compensation (C4)”, has been developed. The model reproduces the differences in oxidation kinetics and hydrogen pickup between different zirconium alloys, such as Zr-Nb and Zircaloy-4. Since oxidized solute elements affect the corrosion process through a space-charge compensation mechanism, synchrotron nano-beam X-ray Absorption Near-Edge Spectroscopy has been performed on multiple oxidized Zr-Nb alloys to determine the oxidation-state profile of niobium in the oxide layer. The results inform the C4 model and the associated hydrogen pickup fraction.

The mechanical properties of zirconiumZirconium alloys are affected by the presence of hydrides. The strain fields around hydridesHydrides , which are affected by the size, orientation, and hydride phase, are believed to influence the apparent hysteresis between solubility limits on heating and cooling. TEM characterization of dislocation fields near hydrides in Zircaloy-4 specimens, which were exposed to 300 °C primary-water conditions for 600 h, was performed both before and after a heating and cooling cycle. In addition, EELS characterization is provided before heating. In situ TEM imaging/recording and nano-diffraction allowed monitoring of the morphology of dissolving hydrides throughout the temperature cycling. No dislocations in the matrix surrounding the hydrides were visible prior to heating; however, when the hydrides dissolved, dislocations were visible in the space the hydrides had previously occupied, providing a map of the original hydride distribution. These dislocation ‘nests’ are likely the preferential sites for subsequent hydride precipitation and elucidate the so-called ‘memory effect’. Advancing the understanding of hydride formation kinetics, hydride morphology, and hydrogen solid solubility limits can help to reduce uncertainties and conservatism when addressing the risks of hydrogen embrittlementHydrogen embrittlement and hydride cracking in zirconium components.

The presence of hydrogen (H) can detrimentally affect the mechanical properties of many metals and alloys. To mitigate these detrimental effects requires fundamental understanding of the thermodynamics and kinetics governing H pickup and hydride formation. In this work, we focus on H diffusion in Zr-based alloys by studying the effects of alloying elements and stress, factors that have been shown to strongly affect H pickup and hydride formation in nuclear fuel claddings. A recently developed accelerated kinetic Monte Carlo method is used for the study. It is found that for the alloys considered here, H diffusivity depends weakly on composition, with negligible effect at high temperatures in the range of 600–1200 K. Therefore, the small variation in H diffusivity caused by variations in compositions of these alloys is likely not a major cause of the very different H pickup rates. In contrast, stress strongly affects H diffusivity. This effect needs to be considered for studying hydride formation and delayed hydride cracking.

Thermal degradation of cast stainless steels was studied to provide an extensive knowledgebase for the assessment of structural integrity during extended operations of reactor coolant systems. The CF3 and CF8 series cast stainless steels with relatively low (5–12%) δ-ferrite contents were thermally aged at 290–400 °C for up to 10,000 h and tested to measure changes in tensile and impact properties. The aging treatments caused significant reduction of tensile ductility, but only slight softening or negligible strength change. The thermal agingThermal aging also caused significant reduction of upper shelf energy and large shift of ductile-brittle transition temperature (ΔDBTT). The most influential factor in thermal degradation was ferrite content because of the major degradation mechanism occurring in the phase, while the nitrogen and carbon contents caused only weak effects. An integrated model is being developed to correlate the mechanical propertyMechanical properties changes with microstructural and compositional parameters.

Cast austenitic stainless steels (CASS)Cast Austenitic Stainless Steels (CASS) consist of a dual-phase microstructure of delta ferrite and austenite. The ferrite phase is critical for the service performance of CASS alloys, but can also undergo significant microstructural changes at elevated temperatures, leading to severe embrittlement. To understand thermal aging embrittlementThermal aging embrittlement , fracture toughnessFracture toughness J-R curve tests were performed on unaged and aged CF8 specimens at 315 ℃. The microstructure of CF8 was also examined before and after thermal aging with transmission electron microscopy and atom probe tomography. While no microstructural change was observed in the austenite after thermal aging, a high density of G-phase precipitates and a phase separation of alpha/alpha prime were detected in ferrite. To study the deformation behavior, tensile tests were performed at room temperature with in situ wide-angle X-ray scatteringWide-Angle X-ray Scattering (WAXS) measurements. The differences in lattice strains between ferrite and austenite were much higher in the aged than in the unaged samples, suggesting a higher degree of incompatible deformation between ferrite and austenite in the aged samples.

Thermal aging degradationThermal aging degradation of cast austenitic stainless steels (CASS) was studied by electron microscopy to understand the mechanisms for thermal embrittlement potentially experienced during extended operations of light water reactor coolant systems. Four CASS alloys—CF3, CF3M, CF8, and CF8M—were thermally aged up 1500 h at 330 and 400 °C, and the microstructural evolution of the material was characterized by analytical aberration-corrected scanning transmission electron microscopy. The primary microstructural and compositional changes during thermal aging were spinodal decompositionSpinodal decomposition of the δ-ferrite into α/α′, precipitation of G-phaseG-phase precipitation in the δ-ferrite, segregation of soluteSolute segregation to the austenite/ferrite interphase boundary, and growth of M23C6 carbides on the austenite/ferrite interphase boundary. These changes were shown to be highly dependent on aging temperature and chemical composition, particularly the amount of C and Mo. A comprehensive model is being developed to correlate the microstructural evolution with mechanical behavior and simulation.

An austenitic stainless steel Type 316L weld was thermally aged for 20,000 h at 400 °C and electrochemical characterization was performed to measure corrosion resistance in δ–ferrite phase. It is well known that a severe thermal aging causes decrease of fracture resistance and increase of the hardness of δ–ferrite, which was related to the spinodal decomposition. After thermal aging, the DL-EPR response of 316L weld was dominated by parent austenite matrix without reactivation peak. To characterize the δ–ferrite only, austenite phase was selectively dissolved from the matrix by electrochemical etching method. The double–loop electrochemical potentiokinetic reactivation (DL-EPR) analysis of the δ–ferrite phase showed degradation in corrosion resistance after thermal aging with the appearance of a cathodic loop and reactivation peak during the reverse scan. The degradation in corrosion resistance of δ–ferrite phase could be attributed to the localized Cr-depletion due to spinodal decomposition and precipitation of intermetallic phases during thermal aging.

The objective of this study was to clarify the effect of long-term thermal aging on SCC initiation susceptibility in low carbon austenitic stainless steels. Specimens used were Type 304L and 316L austenitic stainless steels. Both steels were cold worked to 20% thickness reduction (CW) followed by long-term thermal aging at 288 °C for 14,000 h (LTA). Creviced Bent Beam (CBB) testing was carried out to estimate the SCC initiation susceptibility under BWR simulated water condition at high temperature. The results of the CBB tests showed that Type 304L specimens with CW and LTA treatment exhibited no SCC susceptibility. In contrast, the SCC initiation susceptibility of Type 316L increased by the combination of cold work and long-term thermal aging. To understand these results, evaluations on the changes of microchemistry, microstructure and mechanical properties induced by the CW and LTA treatment have been performed, and their correlation with the SCC initiation susceptibility was discussed.

Cast austenitic stainless steels (CASS)Cast Austenitic Stainless Steels (CASS) used in reactor core internals are subject to high-temperature coolant and energetic neutron irradiationNeutron irradiation during power operations. Due to both thermal agingThermal aging and irradiation embrittlementIrradiation embrittlement , the long-term performance of CASS materials is of concern. To assess the cracking behavior of irradiated CASS alloys, crack growth rate (CGR) and fracture toughness J-R curve tests were performed on two CF3 alloys. Miniature compact tension specimens were irradiated to ~3 dpa, and were tested at ~315 °C in simulated LWR coolant environments with low corrosion potentials. No elevated cracking susceptibility was observed at this dose in the test environments. The power exponents of the 3 dpa J-R curves were much lower than that of unirradiated or irradiated specimens at lower doses, indicating a significant decline in fracture resistance. A preliminary microstructural study revealed irradiation-induced microstructural changes in both austenite and ferrite, suggesting an embrittlement mechanism involving both phases at this dose level.

The integrity of reactor internal components made of austenitic stainless steel weldsAustenitic stainless steel weld with a duplex structure can potentially be affected by thermal agingThermal aging and/or neutron irradiationNeutron irradiation induced embrittlement. There have not been sufficient studies on the long-term service performance of SS welds in light water reactors. In this study, thermal aging was performed at 400 °C for up to 2220 h on a 308L weld, and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 1019 n/cm2, E > 1 meV). The microstructural evolution of the ferrite phase was characterized using atom probe tomography (APT) and auxiliary transmission electron microscope studies. Spinodal decomposition and Ni-Mn-Si solute clusters were observed in both the thermally aged and neutron irradiated 308L welds. As compared with thermal aging, low dose neutron irradiation induced similar spinodal decomposition with slightly larger concentration wavelength and amplitude. The solute clusters in irradiated ferrite phase also show a larger mean size, a wider size distribution, but a lower number density as compared with those in thermally aged ferrite phase. In addition, the neutron irradiation significantly promotes segregation of trace elements, particularly phosphorus, at the Ni-Mn-Si solute clusters.

The combined effect of thermal agingThermal aging and irradiation on cast and welded stainless steelStainless steel solidification structures is not sufficiently investigated. From theory and consecutive aging and irradiation experiments, the effect of simultaneous low rate irradiation and thermal aging is expected to accelerate and modify the aging processes of the ferriteWeld ferrite phase. Here, a detailed analysis of long-term aged material at very low fast neutron flux at LWR operating temperatures using Atom Probe Tomography is presented. Samples of weld material from various positions in the core barrel of the Zorita PWR are examined. The welds have been exposed to 280–285 °C for 38 years at three different neutron fluxes between 1 × 10−5 and 7 × 10−7 dpa/h to a total dose of 0.15–2 dpa. The aging of the ferrite phase occurs by spinodal decompositionSpinodal decomposition , clustering and precipitation of e.g. G-phase. These phenomena are characterized and quantitatively analyzed in order to understand the effect of flux in combination with thermal aging.

A better understanding of stress corrosion cracking (SCC) initiation is one of the keys towards developing proactive mitigation techniques for the safe and economic operation of nuclear power plants. However, SCC initiation laboratory studies are very time consuming and require multiple specimens, Hence, in the framework of the European “MICRIN+” research project, an accelerated test method was evaluated for screening the SCC susceptibility in a relatively short time frame. The effects of surface roughness and strain rate on SCC initiation susceptibility in Alloy 182 weld metal were evaluated in simulated BWR and PWR environments. Constant extension rate tensile tests were performed using flat tapered tensile specimens with different surface finishes (ground and polished) in hydrogenated water at 288 and 340 °C. The surface crack distribution and crack length as well as stress thresholds for SCC initiation were analyzed by detailed post-test quantitative characterization. Some test data were analyzed by the EngInit SCC initiation model. The accelerated test technique was successfully applied and revealed very promising results. The highest crack density and lowest stress thresholds for crack initiation were found on the ground surfaces and at the lowest strain rates. A test’s load response can be fed to the EngInit model; with parameters to be determined by comparing EngInit’s damage to the experimental surface crack distributions. EngInit can potentially be used to link laboratory test results on flat tapered specimens to SCC initiation in components in the field.

Determination of the fracture toughness properties and thermal aging behavior ofDissimilar Metal Weld (DMW) dissimilar metal weld (DMW) joints is of utmost importance for successful structural integrity and lifetime analyses. This paper presents results from fracture resistance (J-R), fracture toughness (T0) and Charpy-V impact toughness tests as well as fractography performed for an industrially manufactured narrow-gap DMW mock-up (SA508-Alloy 52-AISI 316L). Tests were performed on post-weld heat treated, 5000 h aged and 10,000 h aged material. The results show that this DMW is tough at the SA 508-Alloy 52 interface, which typically is the weakest zone of a DMW. The DMW joint maintains its high fracture resistance also after thermal agingAging . Crack propagates for a large part in the carbon-depleted zone (CDZ) of SA 508 but deflects occasionally to the Alloy 52 side due to small weld defects in µm scale. Ductile-to-brittle transition temperature determined from Charpy-V impact toughness tests increases due to thermal aging, but only to a minor extent. No significant change is observed for the T0 transition temperature due to aging.

The properties of oxide filmOxide film formed on stainless steel (SS) cladding on low alloy steel (LAS)Low Alloy Steel (LAS) after immersion in simulated PWR primary water environments with different dissolved oxygen contents are investigated. The HAZ in the LAS consist of overheated crystal region, complete recrystallized region and incompletely recrystallized region, while SS cladding consist of austenite zone and austenite and ferrite mixing zone. Pitting appeared on 309L SS after immersion in high temperature waterHigh temperature water due to the dissolution of inclusions existed previously on 309L SS which contain higher ferrite content. Raman spectra and TEM results show that the outer layer is mainly Fe-rich spinel oxides while the inner layer is mainly Cr-rich oxides. Ni is mainly concentrate at the oxide/substrate interface due to the low oxygen affinity. The inner oxide layer on 308L SS is thinner than that on 309L SS, implying that ferrite distributed on austenite is not favorable for the growth of oxides. Reducing the oxygen content in PWR primary water favored the formation of spinel oxides.

The safe-end dissimilar metal weld (DMW)Dissimilar Metal Weld (DMW) joining the reactor pressure vessel to the main coolant piping is one of the most critical DMWs in a nuclear power plant (NPP). DMWs have varying microstructures at a short distance across the ferritic-austenitic fusion boundary (FB) region. This microstructural variation affects the mechanical properties and fracture behavior and may evolve as a result of thermal aging during long-term operation of an NPP. This paper presents microstructural characterizationMicrostructural characterization performed for as-manufactured and 5000 h and 10,000 h thermally aged narrow-gap DMW representing a safe-end DMW of a modern pressurized water reactor (PWR) NPP. The most significant result of the study is that the thermal agingAging leads to a significant decrease in a hardness gradient observed across the ferritic-austenitic FB of the as-manufactured DMW.

Because a large time spread in most crack initiation tests makes it a daunting task to predict the initiation time of cracking, a probabilistic model, such as the Weibull distribution, has been usually employed to model it. In this case, although it might be anticipated to develop a more reliable cracking model under ideal cracking test conditions (e.g., large number of specimen, narrow censoring interval, etc.), it is not straightforward to quantitatively assess the effects of these experimental conditions on model estimation uncertainty . Therefore, we studied the effects of some key experimental conditions on estimation uncertainties of the Weibull parameters through the Monte Carlo simulations. Simulation results suggested that the estimated scale parameter would be more reliable than the estimated shape parameter from the tests. It was also shown that increasing the number of specimen would be more efficient to reduce the uncertainty of estimators than the more frequent censoring.

This paper covers the results of laboratory examinations performed on a leaking letdown coolerLetdown cooler from Oconee Unit 3. The laboratory scope included dewatering, pressure testing, visual inspections, metallography, Vickers micro-hardness, scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), X-Ray Diffraction (XRD), and Optical Emission Spectroscopy (OES). The laboratory examinations identified one tube containing a through-wall crack. The most likely cause of the crack appeared to be OD-initiated caustic stress corrosion cracking (SCC)Caustic stress corrosion cracking . The presence of heavy deposits on the tube OD surface and heat tinting on the primary and secondary flow seals indicated boiling occurred near the tight radius region of the bundle. Once boiling occurred, caustic-forming species such as calcium phosphate deposited and concentrated on the tube OD surface. The literature indicates as caustic concentrations approach ~20%, the conditions become favorable for caustic SCC to occur in austenitic stainless steels such as Type 316L.

Cracks Crack were found in the Alloy 182Alloy 182 weld and buttering of two core shroudCore shroud support legsSupport leg in the Forsmark BWRBWR Unit 1. This paper presents the root cause analysis based on mechanical and metallurgical analysis by Microscopy, Electron Backscatter Diffraction, and micro- and nano-indentation analysis along the crack path. The High Resolution Analytical Transmission Electron Microscopy examinations on the oxides are presented in detail in another paper at this conference. The morphology of the cracks, local hardness, crystallographic orientations and compositions along the crack path and crack tip regions were examined. The crack penetrated the boat keel surface into the A182 Low Alloy Steel mixing zone. Cracking was identified as IGSCC and signs of grinding and elevated micro-hardness near the boat top surface region were detected. Plastic deformation was observed in a crack tip region in a sample section with the largest crack depth, while not at the crack periphery.

Firefighting water systems are important safety systems in all industries, including nuclear power plants (NPPs). However, they are susceptible to microbially induced corrosion, which is a degradation mode needing special attention. Leakages were observed in a fire fighting system made from stainless steel at a nuclear power plant shortly after maintenance and modernization work, which included replacement of part of the old carbon steel pipelines with stainless steel pipelines, as well as exchange of some Type 304 stainless steel pipes with Type 316 pipes due to relining parts of the system. The failure analysis revealed sub-surface corrosion cavities with pinholes at the inner surface and finally penetrating the whole pipe wall thickness. It was concluded that the reason for the leaks was due to microbially induced corrosion, (MIC). The paper will present the results from failure analyses, explain the remedial actions taken at the power plant, and discuss the implication of these findings on new similar systems, including the importance of avoiding iron deposits and optimization of water quality.

Safety relevant cooling water pipingCooling water piping is designed to transport excess heat both from the reactor core and the fuel pool under all operational and emergency conditions. Plain carbon steels are used, with or without different types of protective coating systems. Outside of the reactor building, such piping is often buried or concealed, which also gives additional protection against damage. Relevant ageing mechanisms are shallow pit corrosion and microbiological induced corrosion (MIC). While these ageing mechanisms can lead to localized leaks and limited loss of cooling water, neither the mechanical integrity nor the cooling ability of the systems are compromised. The extent and ageing management of the mechanisms are described, based on more than 30 years of operating experience. This includes NDT results and post service examinations after retrofits. Based on this positive operational experience, two different NDT concepts are recommended to manage the long time operational ageing.

Stress corrosion cracking (SCC) of stainless steelStainless steel was originally considered only an issue with boiling water reactors (BWRs), but operating experience has shown that this phenomenon also occurs in pressurized water reactors (PWRs), such as in off-chemistry locations of stagnant branch connection piping. In this paper, the safety significant stainless steel piping locations susceptible to SCC are identified for three representative PWR plants (Plant A [Babcock and Wilcox-designed], Plant B [Westinghouse-designed], and Plant C [Combustion Engineering-designed]). For the purpose of this paper, “safety significant” is defined as having a high consequence of failure as determined by the plant’s risk-informed in-service inspection (RI-ISI) program. Weld locations are considered susceptible to SCC when the water is stagnant and ≥200 °F during steady-state reactor operation. The results of this work will be used to develop guidance for selection of welds to inspect when addressing currently existing inspection requirements.

Because atom probe tomography (APT)Atom Probe Tomography (APT) provides three-dimensional reconstructions of small volumes by resolving atomic chemical identities and positions, it is uniquely suited to analyze solute clustering phenomena in materials. A number of approaches have been developed to extract clustering information from the 3D reconstructed dataset, and numerous reports can be found applying these methods to a wide variety of materials questions. However, results from clustering analyses can differ significantly from one report to another, even when performed on similar microstructures, raising questions about the reliability of APT to quantitatively describe solute clustering. In addition, analysis details are often not provided, preventing independent confirmation of the results. With the number of APT research groups growing quickly, the APT community recognizes the need for educating new users about common methods and artefacts, and for developing analysis and data reporting protocols that address issues of reproducibility, errors, and variability. To this end, a round robin experiment was organized among ten different international institutions. The goal is to provide a consistent framework for the analysis of irradiated stainless steels using APT. Through the development of more reliable and reproducible data analysis and through communication, this project also aims to advance the understanding between irradiated microstructure and materials performance by providing more complete quantitative microstructural input for modeling. The results, methods, and findings of this round robin will also apply to other clustering phenomena studied using APT, beyond the theme of radiation damage.

Achieving a better understanding of the Irradiation Assisted Stress Corrosion Cracking resistance is one of the issues to improve the durability of Pressurized Water Reactors. To do so, assessing the interaction of irradiation defects with oxidation of internal vessel bolts, made of 316L alloy, is crucial. In this work we studied the effect of protons pre-irradiations at 1 dpa on the very first steps of oxidation (1 min < t < 24 h) in simulated PWR environment. The morphology of the oxide layer was investigated using optical microscopy and Scanning Electron Microscopy. The oxidation kinetics for short term oxidation is discussed based on the obtained results. It was observed that crystallographic orientation has an effect on the oxidation process. The level of cold-work and the presence of precipitates were taken into account and both seemed to accelerate the oxidation kinetic. Finally, irradiation also tended to speed-up the oxidation phenomenon.

The irradiation-induced defects in stainless steel internal components of pressurized water reactors combined with hydrogen uptake during the oxidation process could be a key parameter in the mechanism for Irradiation-Assisted Stress Corrosion Cracking (IASCC). The ultimate aim of this study is to characterize the effects of irradiation defects on hydrogen uptake during the oxidation of an austenitic stainless steel (SS) in primary water. The focus was made on the interactions between hydrogen and these defects. A heat-treated 316L SS containing a low amount of defects is compared with ion implanted samples. Both materials were characterized by Transmission Electron Microscopy (TEM). Hydrogen uptake was then promoted by cathodic charging using deuterium as isotopic tracer for hydrogen. The deuterium distribution was first characterized by SIMS (Secondary Ion Mass Spectrometry) profiles. This technique highlighted some deuterium segregation in link with the localization of implantation-induced defects, i.e. dislocation loops and cavities. Using TDS (Thermal Desorption Spectrometry) experimental results and literature data, a numerical model was used to simulate the deuterium profiles, providing diffusion and trapping/detrapping information associated with irradiation defects in the 316L SS.

To elucidate the mechanisms of irradiation assisted stress corrosion cracking (IASCC), stress corrosion cracking (SCC) tests on 3 dpa, 19 dpa and 73 dpa neutron-irradiated 316 stainless steel were performed and the effects of irradiation on grain boundary (GB) oxidation were investigated. O-ring specimens were prepared from irradiated flux thimble tubes and a constant load SCC test was performed in a simulated pressurized water reactor primary water at 320 °C. After the SCC test, the oxidation condition of GBs was examined by transmission electron microscopy. Evidence of GB oxidation was found in all examined GBs, even at the relatively low dose of 3 dpa. The morphology of GB oxidation was sharp wedge-shaped. The average GB oxidation length at 3 dpa, 19 dpa and 73 dpa were 100 nm, 340 nm and 400 nm, respectively, indicating the promotion of GB oxidation due to irradiation. In the GB oxide, Fe and Ni depletion and Cr enrichment were observed. Also, Ni enrichment on GB was observed in front of the GB oxidation.

Nickel-base alloys, 625DA and 625Plus have received renewed interest as potential structural materials in nuclear reactors to replace the austenitic stainless steels, which show high susceptibility of irradiation-assisted stress corrosion cracking (IASCC). We investigated the microstructural response of both alloys after 2 MeV protons irradiated to 5dpa at 360 °C in the Michigan Ion Beam Laboratory (MIBL). Transmission electron microscopy was performed on plan-viewed samples with a depth range 9–12 μm prepared by jet-polishing. Detailed analysis included changes in phases, dislocation loops, voids swelling, and radiation induced segregation (RIS). Nano-scaled irradiation-induced precipitates and dislocation loops were pervasive. Voids were absent in these alloys. RIS occurred at random high angle grain boundaries examined. A complete characterization of the irradiated microstructure is required to understand their mechanical and IASCC behavior.

Radiation-induced microstructural changes control the Irradiation Assisted Stress Corrosion Cracking (IASCC)Irradiation-assisted stress corrosion cracking of core materials, which is a key factor in the extension of the operating lifetime of Light Water Reactors (LWRs). Nickel-base alloys are considered as potential structural materials to replace highly IASCC susceptible austenitic stainless steels. Constant extension rate tensile (CERT) tests were conducted on proton irradiated high strength nickel-base alloy 625 with two different heat treatment conditions (625Plus and 625DA) in both simulated BWR NWC and PWR primary water. Crack length per unit area and fraction of grain boundaries that cracked were used to assess the IASCC susceptibility. Both 625Plus and 625DA showed a very high IASCC susceptibility. 625DA also exhibited greater changes in all microstructure features than 625Plus.

A commercial purity 304SS was irradiated to 5 dpa Kinchin-Pease (10 dpa full-cascade) using 2 meV protons at 360 °C. Post-irradiation annealing (PIA) was applied to reduce or remove IASCCIrradiation-Assisted Stress Corrosion Cracking (IASCC) susceptibility. This paper focuses on the links between irradiation-induced hardening and irradiated microstructures of the as-irradiated and PIA conditions; the irradiated microstructure is assessed by transmission electron microscopy (TEM) and atom probe tomography (APT)Atom Probe Tomography (APT) . Dislocation loops, Ni–Si clusters, and Cu-enriched clusters are present in the as-irradiated condition. When the dislocation loopsDislocation loops are removed by PIA, ~40% of the as-irradiated hardnessHardness remains and can be rationally attributed to the solute clustersSolute clusters still present in the PIA microstructure. The observations indicate that hardening in the as-irradiated condition is controlled by both dislocation loops and solute clusters and suggest that radiation-induced solute clusters may be important to detailed understanding of IASCC (irradiation-assisted stress corrosion cracking)Irradiation-Assisted Stress Corrosion Cracking (IASCC) .

The Irradiation-Assisted Stress Corrosion Cracking (IASCC)Irradiation-Assisted Stress Corrosion Cracking (IASCC) initiationInitiation data for austenitic stainless steelsStainless steel in Pressurized Water Reactor (PWR)Pressurized Water Reactor (PWR) primary water environments were collected from available research programs and evaluated by the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The objective was to determine the relationship between applied tensile stress, neutron fluence, and initiation of IASCC at nominally constant load. Analysis of the available data shows that the applied tensile stress level for initiation of IASCC decreases with increasing neutron dose in PWR environments above the PWR threshold for IASCC of three displacements per atom (dpa). An apparent asymptotic value between approximately 30 and 35% of irradiated yield strength has been observed for neutron dose levels between approximately 10 and 100 dpa. Maximum testing times up to approximately 5000 h are now available, but these still are several orders of magnitude less than 60–80 year operating times. However, the results from this study can currently be used by the nuclear industry to assess the effects of irradiation on austenitic stainless steels in PWR systems as an indicator of the combination of stress and neutron dose at which IASCC becomes possible, particularly for subsequent license renewal (SLR) evaluations.

A process for quantitatively measuring the residual stress near dislocation channel—grain boundary interaction sites has been developed especially for irradiated stainless steel using High Resolution Electron Backscatter Diffraction (HREBSD)High resolution electron backscatter diffraction . Tensile stress acting normal to the grain boundary at 15 different discontinuous channel—grain boundary sites were observed to be highly elevated, with peak stresses reaching ~2 GPa, which is roughly an order of magnitude greater than the stresses observed at sites where slip transfer occurred at the grain boundary. A clearly observable difference can be made between the stress profiles present at discontinuous and continuous channel—grain boundary interaction sites. This difference is consistent with the theory that high tensile stress at the grain boundary may be a key driving factor for the initiation of irradiation assisted stress corrosion cracksIrradiation assisted stress corrosion cracking .

Post-irradiation annealing (PIA)Post-irradiation annealing was conducted to investigate the cause of irradiation-assisted stress corrosion cracking (IASCC)IASCC . The effects of PIA on irradiation hardeningHardening , dislocation channelDislocation Channels formation, and IASCC susceptibility were examined for a 304L stainless steelStainless steel irradiated to 5.9 dpa in the Barsebäck-1 reactor (Sweden). The annealing treatments were performed at temperatures in the range 450–600 °C and times ranging from 1–20 h. Longer annealing times and higher temperatures, as represented by iron diffusion distance, resulted in a significant reduction in irradiation hardening. IASCC susceptibility was measured for the as-irradiated and two PIA conditions (500 °C: 1 h and 550 °C: 20 h) via interrupted CERT tests under simulated BWR-NWC conditions. The annealing treatments progressively reduced IASCC susceptibility (as measured by the final intergranular fracture fraction) and dislocation channel density.

The effect of irradiation on stress corrosion cracking susceptibility and fracture toughness of alloy X-750 has been investigated. The material has been irradiated at a target temperature of 288 ℃ in the Advanced Test Reactor at Idaho National Laboratory to a fluence of approximately 1.93 × 1020 n/cm2 (E >1 meV). Stress corrosion cracking crack growth rates were determined in both unirradiated and irradiated materials in normal water chemistry and hydrogen water chemistry environments. Although the effect of irradiation on tensile properties and fracture toughness was observed, there was no significant effect of irradiation observed on the propagation rate of stress corrosion cracks.

In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth rates (CGRs), locally deformed microstructure, and oxide film properties for neutron-irradiated austenitic stainless steels (SSs) are investigated. Crack growth rate tests have been performed in simulated Boiling Water Reactor (BWR) water conditions (at ~288 ℃) on a neutron-irradiated 316L SS at ~12–14 dpa. After the crack growth rate tests, the microstructures near the crack tip of the CT specimens are examined with scanning transmission electron microscope (FE-STEM). In comparison with a previous study at <~2 dpa, the irradiated 316L SS at ~12 dpa shows a less benefit of low electrochemical corrosion potential (ECP) conditions on CGR. An inactive crack tip immersed over 1000 h was filled with oxides, while almost no oxide film was observed near the active crack front in the low-ECP conditions. In addition, a high density of deformation twins and dislocations were found near the fracture surface of the crack front. It is considered that both localized deformation and oxidation are possible dominant factors for the SCC crack growth in highly irradiated SSs.

Crack growth rate (CGR)Crack Growth Rate (CGR) testing in BWR normal water chemistry was performed on compact tension (CT) specimens of two different sizes (B = 8 mm and B = 19 mm), machined from a Type 304 SS core shroud at a dose of ~1 dpa. The objectives were to study the effect of specimen sizeSpecimen size on the CGR, and to determine the K validityK validity limit for a CT specimen dimension used in previous studies. The results show that for materials with significant strain hardening capacity remaining, there is no effect of specimen size on the CGR when testing is conducted at stress intensity factors valid according to ASTM E399 using the flow strength. For materials at higher dose in which the strain hardening capacity is lost or greatly reduced, a different K validity criterion might be applicable.

During stress corrosion crack propagation, stress values in the crack tip vicinity often exceed material yield stress limit. Plastic deformation processes may accompany and influence cracking. Here, stress corrosion crack propagation and deformation mechanisms were investigated using EBSD analysis. The investigated material was 304L Ti-enriched austenitic stainless steel irradiated to 10.4 dpa at 320°C in the BOR-60 fast reactor. Crack growth tests were conducted in a simulated Normal Water Chemistry (NWC) environment in the temperature range 288–320 °C using compact tension specimens. By analyzing crack trajectory and grain structure in the crack vicinity, it was established that grain orientation with respect to the acting stress direction was not a key factor controlling crack propagation. No crystallographic orientation susceptible to cracking was identified. Also, EBSD analysis revealed strong inhomogeneity in plastic strain distribution along the crack path. Most crack-adjacent grains remained virtually strain-free whereas few grains experienced strong plastic strain. These areas were presumed to be “plastic bridges” or “ductile ligaments.”

Simultaneous exposure of 316L stainless steel to a proton beam and high purity water containing 3 wppm H2 was used to study the effect of radiation on corrosion. Protons create displacement damage in the solid and radiolysis products in the water. Irradiations lasted 24 h at a damage rate of 7 × 10−7 dpa/s. The 316L was solution annealed at 1050 °C for 30 min, 5% cold worked, and heat treated at 1100 °C for 10 min resulting in 23 μm grains. Samples were pre-characterized by EBSD to correlate grain orientation with oxide properties measured by Raman spectroscopy and TEM. Following irradiation, hematite was identified exclusively in areas exposed to radiolyzed water, both under the beam and downstream. Inner oxide layers in the unirradiated region had a strong dependence on grain orientation, whereas the irradiated region has little to no grain orientation dependence.

Corrosion Related Unidentified Deposition (CRUD)Corrosion Related Unidentified Deposition (CRUD) in PWR may cause severe issues, such as Tube Support Plate (TSP) blockage, fuel cladding cracking, and subsequently increased radiation doses for workers. The primary objective of this work is to develop an all-inclusive deposition model, which will reproduce the morphology and elucidate the contributing electrokinetic mechanisms. In this paper the development and verification of a model of the streaming current linking the potential distribution and the fluid flow behaviour using the Finite Element Method (FEM)Finite Element Method (FEM) is presented. In the model, coupled anodic and cathodic regions were found at the inlet of a pipe restriction, associated with a region of recirculating flow following the front facing step (FFS). The corresponding current densities and overpotential at the metal/solution interface were calculated. The coupled anode and cathode may explain the observed deposition process—generating deposits at the front facing step first, followed by a region free of deposits and then repeating ripples of deposited material. At the restriction outlet, a cathode was found which balances the current loops. In this paper, the simulated initiation and propagation processes of the electrokinetic depositionElectrokinetic deposition are presented.

The iron contentIron content in Ni–Cr–xFe (x = 0–9 at.%) alloys strongly affected the properties of oxide films Oxide film formed in a simulated PWR primary water environment at 310 °C. Increasing the iron content in the alloys increased the amount of iron-bearing polyhedral spinel oxide particles in the outer oxide layer and facilitated the local oxidation penetrationLocal oxidation penetration into the alloy matrix from the chromium-rich inner oxide layer. The local oxidation penetration was caused by the pile-up of the cation vacancies.

Alloy 690TTAlloy 690TT has been shown to be susceptible to leadLead stress corrosion cracking (PbSCC)PbSCC at pHT values of 9.1 and above and with no cracks occurring at lower pHT values like 8.5. Previous work on PbSCC has been completed at the open circuit potential (OCP). A test program has been completed at applied electrochemical potentials for Alloy 690TT at pHT values of 8.5 and 9.5. Testing has shown sporadic PbSCC occurrence at pHT 8.5 which previously showed no cracking although an exact causal factor was not identified. At pHT 9.5, applying a potential did not stop PbSCC from occurring, although at +75 mV applied potential there was the inclusion of an apparent incubation time where none had previously been observed. To minimize to PbSCC, scoping testing was completed on four candidate inhibitorsInhibitor at pHT 9.5. Three of the four inhibitors tested (TiO2, H3BO3, and CeB6) showed reductions in maximum crack depths compared to testing without an inhibitor. Although promising, inhibitors still require additional testing before widespread use can be recommended.

In spite of strong industry improvements (new alloys, more stringent chemistry control…), Outer Diameter Stress Corrosion Cracking (ODSCC) still occurs in modern SG alloys, although at a lesser extent than it used to affect 600MA tubesTubes . Currently, alloy 600TT, and even 800NG, do indeed suffer from ODSCC. Since the chemistry of the secondary sideSecondary side of PWR NPPs improved, past tests do not enable to assess simply the risks for modern alloys (too extreme pH for instance). Thus, IRSN reviewed what typical chemical conditions could be met in actual steam generatorsSteam generators . Based on this review, IRSN performed corrosionCorrosion tests in these conditions to assess the potential risks for alloys 600TT and 690TT in “typical” top-of-tubesheet environments. This paper will present the safety, technical and industrial frames of these tests, as well as the first results. An emphasis will also be put on the scientific implications of these results.

Alloy 800 has been demonstrated to be susceptible to Pb-induced degradation such as stress corrosion cracking (SCC) in laboratory experiments. PbPb has been proposed to cause such degradation by affecting the formation of protective oxides comprising the passive filmsPassive film on the alloys. To understand the detrimental effects of Pb, Alloy 800 samples pre-passivated under all volatile treatment (AVT) conditions were exposed to alkaline environment of pH280 °C 9.5 at 280 °C in the absence and presence of Pb. The Pb-free and Pb-containing passivated surfaces, along with bare surfaces as control samples, were characterized using electrochemical and surface analytical techniques. The results contrast the susceptibility of Alloy 800Alloy 800 to Pb-induced degradation reported in literature, where experiments are usually performed on bare surfaces with excess Pb, typically as PbO, present in the aqueous solutions; suggesting that the effect of Pb depends on whether it is present in solution or incorporated in the film.

Thermal embrittlement caused by phase transformations in the temperature range of 204–538 °C limits the service temperature of duplex stainless steels. The present study investigates a set of wrought (2003, 2101, and 2205) and weld (2209-w and 2101-w) alloys in order to better understand how alloying elements affect thermal embrittlement. Samples were aged at 427 °C for up to 10,000 h. The embrittlement and thermal instability were assessed via nanoindentation, impact toughness testing, and atom probe tomography (APT). Results demonstrate that the spinodal amplitude is not an accurate predictor of mechanical degradation, and that nanoindentation within the ferrite grains served as a reasonable approximate for the embrittlement behavior. Compositionally, alloys with a lower concentration of Cr, Mo, and Ni were found to exhibit superior mechanical properties following aging.

Alloy 800 nuclear grade (NG) is a material of choice for replacement steam generators (SG) due to its inherent resistance to primary water Stress Corrosion Cracking (SCC) stress corrosion cracking (SCC). However, the long term performance of SGs depends on the performance of the material in upset conditions. Various degradation modes have been observed in Alloy 800NG under simulated secondary crevice environments (SCE) in C-ring and CERT experiments. Furthermore, the first incidences of SCE SCC have been observed in Alloy 800NG SG tubes in nuclear power plants and may be the sentinel events at the onset of more extensive cracking in the future. Understanding the parametric dependencies of SCC obtained under representative SCE and plausible transient conditions are keys to predicting future SG performance, validating mitigation strategies, and addressing life extension issues. The results of SCE crack growth rate (CGR) testing of Alloy 800 Alloy 800NG in conditions representative of an acid-sulfate chemistry upset condition will be presented.

Aggressive aqueous environments (Pb, S, pH extremes) used in laboratory tests have been shown to induce stress corrosion cracking (SCC) in Ni–Fe–Cr alloys. These conditions are used to simulate the extremes of secondary side crevice environments that are unlikely to occur under normal operating conditions but laboratory testing can still be used to establish sensitivities to abnormal chemistry conditions. Advances in modern microscopy have enabled the characterization of these secondary-side SCC systems at near-atomic resolution, helping to reveal mechanistic characteristics unique to each SCC mode. International progress investigating secondary-side SCC phenomena using analytical transmission electron microscopy (TEM)Transmission Electron Microscopy (TEM) is reviewed in this paper. The unique chemistry and degradation associated with different modes of SCC are identified and compared among Ni–Fe–Cr steam generator tube alloys of interest (Alloy 690 and Alloy 800). It is revealed that each SCC mode exhibits distinctive characteristics, or a “fingerprint”, which can be used to identify the aggressive environment responsible for inducing SCC.